ML20086F445

Cycle 14 Reload Evaluation

ML20086F445
Person / Time
Site:	Fort Calhoun
Issue date:	11/27/1991
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20086F416	List: ML20086F415 ML20086F423 ML20086F433 ML20086F445
References
NUDOCS 9112030230
Download: ML20086F445 (62)

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Omaha Public Power District .

Fort Calhoun Station Unit No.1 <

k Cycle 14 Reload Evaluation .

JQhl j@55E2dbb;5kfy;2*

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FORT CALHOUN STATION UNIT NO.1 CYCLE 14 RELOAD EVALUAT!ON TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 O P E RATI N G H IS TO RY O F CYC LE 13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 G E N E R AL D E S C RI PTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.0 FU E L SYST E M D E S I G N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ~ 20 5.0 N U C LEAR D E S IG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1 PHYSICAL CHARACTERISTICS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1.1 Fuel Mana g ement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.1.2 Power Distribution ............... . ................... 22 5.1.3 Safety Related Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.1.3.1 Ejected CEA Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.1.3.2 D ropped C EA Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.2 ANALYTICAL INPUT TO INCORE MEASUREMENTS . . . . . . . . . . . . . . . . 23 5.3 NUCLEAR DESIGN METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.4 - UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS . . . . . . . , . 23 6.0 TH ERMAL- HYDRAUUC DESIG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 6.1 D N B R A N ALYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 6.2 FUEL ROD BOWIN G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 Page 2 of 62

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b FORT CALHOUN STATION UNIT NO.1 CYCLE 14 RELOAD EVALUATION TABLE OF CONTENTS (Continued)

Pagg_

7.0 TRAN SI ENT AN ALYSI S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34  ;

7.1 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGOF3Y 1) ..,, 09 7.1.1 RCS Deproswrization Event . . . . . . . . . . . . . . . . . . . . . . . . . . .. 39 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) . . . . . , 41 7.2.1 Excess Load Event ......................... ......... .. 41 7.2.2 C EA Withdrawal Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 43 7.2.3 Loss of Coolant Flow Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 7.2.4 Full Length CEA Drop Event . . . . . . . . . . . . . . , . . . . . . . . . . . . . . 48 7.2.5 Boron Dilution Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 7.3 P O STU LAT E D ACC I D E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 7.3.1 C EA Ej e ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... 53 7.3.2 Steam Line . Break Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 7.3.3 Seized Rotor Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 8.0 ECCS PERFORMANCE ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 9.0 STARTU P T ESi l N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 10.0 R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 l

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1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation of the design and performanco for the operation of Fort Calhoun Station Unit No.1 during its fourteenth fuel cycle at a full rated power of 1500 MW1. Planned operating conditions iemain the same as those for Cycle 13, unless otherwiso noted in the proposed Coro Operating Limits Report and Technical Specification changes.

The coro will consist of 80 presently operating Batches M, N and P assemblics,52 fresh Batch R assemblics and 1 Batch M assembly discharged from a previous cycle.

The Cyclo '4 analysis is based on a Cyclo 13 termination point betwoon 14,250 MWD /MTU and 15,250 MWD /MTU. In performing analyses of desl0n basis events, limiting sofoty system sottin0s and limiting conditions for operation, limiting values of key parameters were chosen to assure that expected Cycle 14 conditions would be onveloped, provided the Cyclo 13 termination point falls within the above range. The analysis presented herein will accommodato a Cycle 14 length of up to 14,000 MWD /MTU with a coastdown of an additional 1,000 MWD /MTU.

The ovaluation of the reload coro characteristics has been conducted with respect to the Fort Calhoun Station Unit No.1 Cycle 13 safety analysis described in the 1991 updato of the USAR, hereafter referred to as the "referenco cycle" in this report unless noted otherwise.

Specific coro differences have been accounted for in the present analysis. In all cases,it has been concluded that either the referenco cycle analyses envelope the new conditions or the revised analyses presented horcin continuo to show acceptable results. Where dictated by variations from the previous cycle, proposed modifications to the Technical Specifications have boon provided or are being incorporated into the Cycle 14 Coro Operating Limits Report.

The Cycle 14 coro has boon designed to minimize the neutron flux to limiting reactor pressurovesselwoldstoreducothoratoof RTm shiftonthosowelds. Thiswillmaximizo the time to reach the screening criteria that is consistent with the procedure for calculating the amount of radiation embrittlement that a reactor vessel recolves given in Regulatory Guido 1.99, Revision 2 and recently incorporated into 10 CFR 50.61.

The reload analysis presented in this report was performed utilizing the methodology documented in Omaha Public Power District's reload analysis methodology reports (References 1,2, and 3).

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I 2.0 OPERATING HISTORY OF CYCLE 13 Fort Calhoun Station is presently operating in its thirteenth fuel cyclo utilizing Batches L, j M, N and P fuel assemblies. Fort Calhoun Cycle 13 operation began when criticality was achloved on May 25,1990, and full power reached on June 18,1990.1ho reactor has operated up to the present time with the coro reactivity, power distributions, and peaking factors having closely followed the calculateo prodletions.  ;

it is estimated that Cycle 13 will be terminated on or about February 1 1992. The Cycle 1 13 termination point can vary betwoon 14,250 MWD /MTU and 15,250 MWD /MTU ar d i still be within the assumptions of the Cycle 14 analyses. As of November 3,1991, tno  !

Cyclo 13 burnup had reached 12,569 MWD /MTU. l t

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3.0 GENERAL DESCRIPTION The Cycle 14 core will consist of the number and type of assemblies and fuel batches i shown in Tablo 3-1. Eight L assemblies,41 M assemblics and 4 N assemblics will be discharged this outago. They will be replaced by 4 fresh Batch R1 assemblies (0.74 w/o i natural enrichment),10 fresh Batch R2 assemblics (3.85 w/o averago enrichment with '

28 IFBA rods at 0.003 gm Bi. / inch),4 fresh Batch R3 assemblies (3.85 w/o averago enrichment with 48 IFBA rods at 0.003 gm B . / inch),8 fresh Batch R4 assemblies (3.85 w/o average enrichrnent with 64 IFBA rods at 0.003 gm B . / inch),12 fret.h Batch R5 assemblies (3.85 w/o average enrichment with 84 IFBA rods at 0.003 gm Bi / inch),4 fresh Batch R6 assemblies (3.60 w/o averago enrichment with 84 lFBA rods at 0.003 gm B,. / Inch) and 4 fresh Batch R7 assemblies (3.60 w/o average enrichment with 64 IFBA rods at 0.003 gm B , / inch). In addition, the center assembly will be replaced by a Batch M ec 9mbly which was discharged after Cycle 12 and is currnntly residing in Region 1 of the spent fuel pool.

Figure 3-1 shows the fuel management pattern to be employed in Cycle 14. Several changes in fuel management strategy have been incorporated for Cycle 14. First, the overall fuel management scheme is designed to maximize the reduction in neutron leakago seen by the reactor vessel and limiting vessel weld locations. This strategy is called "extramo low radial leakage fuel management" and is very similar to the fuel management previously used in the Cycle 10 core ioading pattern. Listed below aro the specific changes which comprise the extreme low radial leakage fuel managemont strategy:

1) Twelve fuel assemblies on the core periphery will contain four full-length hafnium flux suppression rods por fuel assembly to locally reduco neutron flux near the limiting reactor vessel wolds. Each of the hafnium rods will be placed in one of the outer CEA guido tubes of peripheral fuel assemblies,

2) Four fuel assemblies will contain natural uranium fuel rods which are located on the core periphery adjacent to the reactor vessellimiting welds. These four peripheral assembly locations could not support the use o' full-length hafnium flux suppression rods due to the residence of CEA Shutdown Group A rods.

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! 3)- Use of an integral fuel burnable absorber (IFBA) instead of the traditional fuel l displacing poison rods within selected new fuel assemblies. The lFBA rods consist

of fuel pellets treated with an electrostatically applied, zirconium-diborldo coating

_ which surrounds the fuel pellet circumference. By using IFBA rods, extreme low radial leakage fuel management can provide greater reduction in vessel flux by i

increasing the number of fuel rods available to produce the rated powo. of 1500 l- MWt, thus gaining radial peaking factor margin which is needed to absorb the inward roll of the coro power distribution caused, in part, by the peripheral flux reduction.

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3.0 GENERAL DESCRIPTION (Continued)

The fuel rod and poison rod locations in Batches M and N shimmed assemblies are shown in Figure 3-2. Figuro 3-3 shows the fuel tod locations in Batches N and P unshimmed assemblies. The fuel and poison rod locations for Batch P shimmed ,

assemblies with the fuel rod zone loading technique are shown in Figure 3-4. Due to the Fort Calhoun fuel assembly design, the fuel rods surrounding the fivo largo water holes produce the highest power peaking factors within an assembly. The fuel rod zone loading technioue lowers the initial enrichment of U-235 in those fuel rods while maintaining an assembly overage initial enrichment sufficient to achieve the Cycle 14 design exposure. Figuro 3-5 shows the fuel rod locations for the Batch R1 natural uranium assemblies. Figures 3 -6 through 3-9 provido a diagram of each type of fresh assembly which contains IFBA rods.

The average initial enrichment of tho 52 fresh Batch R assemblics is 3.57 w/o U-235, a reduction of 0.09 w/o from Cycle 13. Excluding the four fresh natural uranium ,

assemblios, the averzgo initial enrichment is 3.81 w/0 U-235. For the second consecutivo cycle, the fuel assembly zone loading technique is used to lower the radial power peaking factors within Batches R2 through R7. Batch R2 through R5 assemblies have fuel rods at both 4.0 w/o enriched U -235 and 3.5 w/o enriched U- 235, while Batch R6 and R7 assemblics have fuel rods at both 3.75 w/o enriched U-235 and 3.25 w/o enriched U-235.

• Figure 3- 10 shows ti.o beginning of Cycle 14 assembly burnup distribution for a Cycle l 13 termination burnup of 15,250 MWD /MTU. The fuel average dischargo exposure at the end of Cycle 13 is projected to be 15,000 MWD /MTU. The initial enrichment of each fuel assembly is also shown in Figuro 3-10. Figure 3- 11 shows the projected end of Cycle 14 assembly burnup distribution. The end of Cycle 14 core averago exposure will be approximately 28,459 MWD /MTU.

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TABLE 3-1 i FORT CALHOUN UNIT NO.1 CYCLE 14 CORE LOADING Initial BOC EOC Poison IFBA Poison Assembly . Burnup** Rods per Rods per Loading Number of Avg.

Designation Assemblies MWD /y70 Burnup* AvgAWD/uru Assembly Assembly gm 0 3ohn.

M/ 1 30,957 45,607 8 -

0.024 N 20 28,485 38,842 0 -

0 N/ 20 31,877 38,303 8 -

0.020 P 8 13,616 30,170 0 --

0 P/ 32 19,256 34,392 8 -

0.027 R1 4 0 4,371 -

0 0.003 R2 16 0 13,096 -

28 0.003 R3 4 0 19,902 -

48 0.003 R4 8 0 19,704 -

64 0.003 R5 12 0 20,739 -

84 0.003 R6 4 0 20,419 -

84 0.003 R7 4 0 -

19,170 -

64 0.003 Assumes EOC13=15,250 uwo/Miu Assumes EOC14=14,000 uwo/MTu Page 8 of 62

I AA - Assembly Location BB - Fuel Type Hf - Location of Hafnium Rods 1 2 N N/

Hf Hf 3 4 5 6 7 N/ R2 P/ R2 P/

8 9 10 11 12 13 N/ R2 P R7 N R3 Hf 14 15 16 17 18 19 R1 P R5 P/ R5 P/

20 21 22 23 24 20 as P/ R4 P/ P/ N/ R6

' N 30 32 27 28 29 31 R2 N R5 N R4 P/

33 N/ 34 35 36 37 38 39 Q P/ R3 P/ R6 P/ M/

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l Cycle 14 Core Loading Pattern Omaha Public Power District Figure Fort Calhoun Station Unit No.1 3-1 Page 9 of 62

00000000000000 08000000000000 OO 000000 OO 00 000000 OO 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00009000000000 OO 000000 OO OO 000000 OO 09000000000000 00000000000000

$ - Shim (B C) 4 Rod (8)

O - Fuel Rod (168)

! - Guide Tube l

l Batches M/ and N/ Assembly Omaha Public Power District Figure Fuel Rod and Poison Rod Locations Fort Calhoun Station Unit No.1 3-2 Page 10 of 62

00000000000000 00000000000000 OO 000000 OO -

OO 000000 OO 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00000000000000 OO 000000 OO OO 000000 OO 00000000000000 00000000000000 O - Fuel Rod (176) 1

- Guide Tube l

Batches N and P Fuel Rod Omaha Public Power District Figure Locations Fort Calhoun Station Unit No.1 3-3 Page 11 of 62

00000000000000 09000000000000 00 000000 00 OO 000000 00 00009000090000 00000000000000 000000 000000 OOOOOO OOOOOO 00000000000000 00009000000000 00 000000 00 OO 000000 00 09000000000000 00000000000000 0 - Shim (B 4C) Rod (8) 0 - Low (3.25 w/o) Enrichment Fuel Rod (88)

O - High (3.95 w/o) Enrichment Fuel Rod (80)

- Guide Tube Batch P/ Assembly Fuel Rod Omaha Public Power District Figure and^* Poison Rod Locations Fort Calhoun Station Unit No.1 3-4 Page 12 of 62

OOOOOOOOOOOOOO OOOOOOOOOOOOOO OO OOOOOO OO OO OOOOOO OO OOOOOOOOOOOOOO OOOOOOOOOOOOOO OOOOOO OOOOOO OOOOOO OOOOOO OOOOOOOOOOOOOO OOO'OOOOOOOOOOO OO OOOOOO OO OO OOOOOO OO OOOOOOOOOOOOOO OOOOOOOOOOOOOO O - Natural Enriched Fuel Rod (176)

- Guide Tube Batch R1 Assembly Fuel Rod Omaha Public Power District Figure Locations Fort Calhoun Station Unit No.-1 3-5 i

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00000000000000  ;

00000000000000  ;

OO 000000 OO l OO 000000 00  ;

00000000000000 00000000000000 000000 000000 000000 ]OOOOOO 00000000000000 00000000000000 OO 000000 GO OO 000000 00 00000000000000 00000000000000 O - Low (3.5 */o) Enrichment Fuel Rod (36)

O - Low (3.5 W/o) Enrichment Fuel Rod with IFBA (16)

O - High (4.0 W/o) Enrichment Fuel Rod (112)

O - High (4.0 W/o) Enrichment Fuel Rod with IFBA (12)

- Guide Tube Batch R2 Assembly Fuel Rod and Omaha Public Power District ' Figure 28 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-6 Page 14 of 62

4 00000000000000 00000000000000 OO 000000 00 OO 000000 OO 00000000000000 00000000000000 000000 000000 .

000000 000000 00000000000000 00000000000000 OO 000000 OO OO 000000 00 00000000000000 00000000000000 O - Low (3,5 W/o) Enrichment Fuel Rod (20)

O - Low (3.5 */o) Enrichment Fuel Rod with IFBA (32)

O - High (4.0 */o) Enrichment Fuel Rod (108)

O - High (4.0 W/o) Enrichment Fuel Rod with IFBA (16)

- Guide Tube

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Batch R3 Assembly Fuel Rod and Omaha Public Power District Figure 48 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-7 Page 15 of 62

00000000000000  ;

00000000000000 00 000000- 00 00 000000 OO 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00000000000000 00 '000000 OO OO OOOOOO OO OOOOOOOOOOO00O -

00OOOOOOOOOOOO O - Low Enrichment Fuel Rod (12) 0 - Low Enrichment Fuel Rod with IFBA (40) 0 - High Enrichment Fuel Rod (100)

O - High Enrichment Fuel Rod with IFBA -(24)

- Guide Tube Batch Enrichnie t (w/o). Enrich n t (w/o) ,

R4 3.5 4.0 R7 3.25 3.75 Batches R4 and R7 Assembly Fuel Omaha Public Power District Figuro Rod and 64 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-8 Page 16 of 62

00000000000000 00000000000000 OO 000000 OO OO .

000000 OO 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00000000000000 00 000000n00 OO OOOOOOVOO 00000000000000 00000000000000 O - Low Enrichment Fuel Rod (12) 0 - Low Enrichment Fuel Rod with IFBA (40)

O - High Enrichment Fuel Rod (80)

O - High Enrichment Fuel Rod with IFBA (44)

- Guide Tube Batch Enrich e$t (w/o) Enricht t (w/o)

R5 3.5 4.0 R6 3.25 3.75 Batches R5 and R6 Assembly Fuel Omaha Public Power District Figure ,

Rod and 84 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-9 l Page 17 of 62

AA - Assembly Location BB - Fuel Type C.CC - Initial Enrichment DD DDD - Assembly Averago(w/o U-235)Exposuro (MWD /MTU)

N " N/

3.70 3.70 24,691 30,556 N/

3.70 R2 3.85

' P/

3.59 R2 3.85 P/

3.59 33,888 0 16,972 0 21,250

" N/ R2 P R7 '"N ' R3 3.70 3.85 3.94 3.60 3.70 3.85 33,896 0 13,618 0 31,088 0 14 15 16 17 18 19 R1 P R5 P/ R5 P/

0.74 3.94 3.85 3.59 3.85 3.59 0 13,615 0 21,003 0 20,941 P/ R4 P/ P/ N/

• R6 26 3.59 3.85 3.59 3.59 3.70 3.60 N 16.964 0 21,006 15.148 30,506 0 3.70 27 28 29 30 31 32 24,691 R2 N R5 N R4 P/

33 3.85 3.70 3.85 3.70 3.85 3.59 0 31,077 0 30,880 0 21,059 N/

3.70 34 35 36 37 38 39 30.540 P/ R3 P/ R6 P/ M/

3.59 3.85 3.59 3.60 3.59 3.80 21,251 0 20,327 0 21,080 30,957 Note: EOC 13 Burnup = 15,250 MWD /MTU Cycle 14 BOC Initial Enrichment Omaha Public Power District Figure and Assembly Average Exposure Fort Calhoun Station Unit No 1 3-10 Page 18 of 62

AA - Assembly Location BB - Fuel Type O.CC - Initial Enrichment (w/o U-r35)

DD DDD - Assembly Average Exposare (MWD /MTU)

'N3.70

• N/

3.70 28,583 33,931 N/ R2 P/ R2 P/

3.70 3.85 3.59 3 85 3.59 38,924 11,964 28,658 14,692 32,892

• N/ R2 P R7 '* N R3 3.70 3.85 3.94 3.60 3.70 3.85 36,825 13,972 30,899 19,170 45,196 19,719 14 15 16 17 18 19 R1 P R5 P/ R5 P/

0.74 &B4 J C5 3.59 3.85 3.59 4

,, J7 - , 29,441 20,891 38,317 20,673 38,469

'20 21 22 23 24 25 26 3.59 3.85 3.59 3.59 3.70 3.60 N 27,5c5 ,8,788 36.16e 32,754 45,878 20,383 3.70 27 IA 29 30 31 32 29.286 RE I'  ; R5 N R4' P/

33 3.8t 3.70 3.85 3.70 3.85 3.59 15,354 45,250 23,652 45,896 20,620 37,929 N/

3.70 34 35 36 37 38 39

, 35,955 , P/ R3 P/ R6 P/ M/

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3.59 3.85 3.59 . 3.60 3.59 3.80 33,981 20.084 38,069 l 20,456 38,007 45,601 1

Cycle 14 EOC Initial Enrichment Omaha Public Power District Figure and Assembly Average Exposure Foit Calhoun Station Unit No.1 3 -11 Page 19 of 62

4.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch R fuelis slightly different from the Batch P fuel due

- to a change to Westinghouse (W) as the fuel vendor,

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The Batch R fuelis similarin design to the fuel supplied by Combustion Engineering and is mochanically, thermally, and hydraulically compatible with the ABB-CE fuel remaining in the Cycle 14 core. References 4 and 5 describe Batches M and R fuel characteristics and design, respectively. The Westinghouse fuel will not be resident in the reactor with any of the Enon (Siemens) fuel previously used at Fort Calhoun.

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5.0 NUCLEAR DESIGN 5.1- PHYSICAL CHARACTERISTICS 5.1.1 Fuel Management The Cycle 14 fuel management uses an extreme low radial leakage design, with twice burned assemblies pradominantly loaded on the periphery of the core with hafnium flux supp'ession rods inserted into the guide tubes of selected peripheral fuel asset nblies adjacent to the reactor vessel limiting welds. This extreme low radial leakage fuel loading pattern is utilized to mir9mize the flux to the pressure vessel welds and achieve the maximum in neutron economy. Use of this type of fuel management 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 14 have been reduced by lowering the enrichment of the f uel pins adjacent to the fuel assembly water holes as described in Section 3.0.

Also described in Section 3.0 is the Cycle 14 loading pattern which is composed of 52 fresh Batch R assemblies of which 48 contain the aforementioned IFBA pellet desigr;. The remaining 4 Batch R assemblies contain fuel rods that are loaded with naturally enriched uranium and also placed in locations near the limiting welds. All of these 48 assemblies-employ intra-assembly uranium enrichment splits. Batches R2 through R5 contain a high pin enrichment of 4.00 w/o and a low pin enrichment of 3.50 w/o, Batches R6 and R7 contain a high pin enrichment of 3.75 w/o and a low pin enrichment of 3.25 w/o. Forty twice burned N assemblies are being returned to the core, along with 40 once burned P assemblies.

One twice bumed M assembly, which was discharged into the spent fuel pool at the end of Cycle 12, will be retumed to the core and used as the center assembly. This assembly arrangement will produce a Cycle 14 loading pattern with a cycle energy of 14,000 MWD /MTU with an additional 1,000 MWD /MTU of energy in a coastdown mode if required.

The Cycle 14 core characteristics have been examined for a Cycle 13 termination-between 14.250 MWD /MTU and 15,250 MWD /MTU and limiting values established for the safety analysis. The Cycle 14 loading pattern is valid for any Cycle 13 end,' ' 5etween these values.

Physics characteristics including reactivity coefficients for Cycle 14 are listed in Table 5-1 along with the corresponding values from Cycle 13. It should be noted that the values of parameters actually employed in the safety analyses are different from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted values with accommodation for appropriate uncertainties and allowances.

The BOC, HZP Main Steam Line Break accident is the most limiting accident of those used in the determination of required shutdown margin Page 21 of 62

5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.1 Fuel Management (Continued) for complianco with Technical Specifications. Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the Cycle 14 BOC, HZP, MSLB -accident. The Cycle 14 values, calculated for minimum scram worth, exceed the minimum value required by Technical Specificatiorss and thus provide an adequate shutdown margin.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrato the all rods ou: (ARO) planar radial power distributions at BOC14, MOC14, and EOC14, respectively, and are based upon the Cycle 13 l ate window burnup timepoint. These racial power densities are assembly averages representative of the entire core length. The high burnup end of the Cyc.ie 13 shutdown window tends to increase the power peaking in the high power assemblies in the Cycle 14 fuel loading pattern. The radial power distributions, with Bank 4 fully inserted at beginning and end of Cycle 14, are shown in Figures 5-4 and 5-5, respectively.

The radial power vistributions described in this section are calculated data without uncertainties or other allowances with the exception of the single rod power peaking values. For both DNB and kW/ft safety and setpoint analyses in either rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time during Cycle 14. 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.

As previously indicated, Figures 3-5 and 3-6 show the integrated assembly bumup values at 0 and 14,000 MWD /MTU, based on an EOC13 burnup of 15,250 MWD /MTU.

The range of allowable axial peaking is defined by the limiting conditions for operation and their axial shape index (ASI). Within these ASI limits, the necessary DNBR and kW/ft margins are maintained for a wide range of possible axial shares. The maximum three-dimensional or total peaking factor (Fn) anticipated in Cycle 14 during normal base load, all rods out operation at full power is 2.095, it :luding uncertainty allowances.

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5.0 @CLE.AR DES!GN (Continued) -

5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data Bounding 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 14. These values are

' projected to encompass the worst conditions anticipated during Cycles 14 through 16. The values shown bound actual Cycle 14 values which were calculated in accordance with Reference 3. In addition, Table 5-3 lists these values used from Cycle 13 for comparison.

3 5.1.3.2 Dropped CEA Data The Cycle 14 safety related data for the dropped CEA analysis were calculated identically with the methods used in Cycle 13.

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 13.

5.3 NUCLEAR DESIGN METHODOLOGY Analyses have been performed in the manner and with the methodologies documented in References 1 and 2.

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties which are applied to Cycle 14 are the same as those presented in Reference 2.

Page 23 of_62-

. - - . . . _ . . - . _ - ~ .. - _- . . - . - - - . . -

TABLE 5- 1 FORT CALHOUN UNIT NO.1, CYCLE 14 NOMINAL PHYSICS CHARACTERISTICS Units .

Cycle 13 Cyc!e 14 Critical Boron Concentration Hot Full Power, ARO, Equilibrium Xenon, BOC ppm 1187 835 Inverse Boron Worth Hot Full Power, BOC ppm /%Ap 112 113 Hot Full Power, EOC ppm /%Ap 84 90 Reactivity Coefficients with All CEAs Withdrawn Moderator Temperature Coefficient (MTC) (includes uncertainties)

Beginning of Cycle, HZP 10 ~'Ap/' F + 0.51 * + 0.09 End of Cycle, HFP 10Ap/ F -2.47 - 2.80 *

• Doppler Coefficient (FTC)

Hot Full Power, BOC 10~'Ap/ F -1.66 - 1.51 Hot Full Power, EOC 10-'Ap/* F -1.85 -1.69 Total Delayed Neutron Fraction, p.,,

BOC 0.00614 0.00625 EOC 0.00519 0.00518 Neutron Generation Time I.

BOC 10 ~'sec 21.6 21.6 EOC 10-'sec 28.8 27.2

. This value exceeds the Technical Specification limit of +0.E O x 10~'Ap/"F, however, the actual MTC at HZP, BOC, including uncertainties, did not exceed tne Technical Specification limit.

• • This value exceeds the current Technical Specification limit of -2.70 x.10Ap/*F, therefore, a chan e to Technical Specification 2.10.2(3)c. is being made to lower the limit to -3.00 x 10 ~' F including uncertainties.

Page 24 of 62

TABLE 5-2 FORT CALHOUN UNIT NO.1.-CYCLE 14 LIMITING VALUES OF REACTIVIT/ WORTHS AND ALLOWANCES FOR HOT ZERO POWER MAIN STEAM LINE BREAK, %Ap Cycle 13 Cycle 14

1. Worth of all CEAs inserted 9.23 7.52

2. Stuck CEA Allowance 1.83 1.17

3. Worth of all CEAs Less Worth of Most Reactive CEA Stuck Out 7.40 6.35

4. Power Dependent insertion Limit CEA Worth 1.23 1.19

5. Calculated Scram Worth 6.17 5.16

6. Physics Uncertainty plus Bias 0.80* 0.10**

7. Net Available Scram Worth 5.37 5.06

8. Technical Specification Shutdown Margin 4.00 4.00

9. Margin in Excess of Technical Specification Shutdown Margin 1.37 1.06 13% of calculated scram worth.

1.96% of calculated scram worth from revised ABB-CE methodology biases and uncertainties.

Page 25 of 62

TABLE 5-3 FORT CALHOUN UNIT NO.1, CYCLE 14 BOUNDING CEA EJECTION DATA Ma.vimum Radial Power Peaking Factor BOC 13 Value EOC 13 Value BOC 14 Value EOC 14 Value Full Power with Bank 4 inserted; worst CEA ejected 2.41 2.68 3.73 3.73 Zero Power with Banks 4+3 inseded; worst CEA ejected 3.43 3.99 5.74 5.74 Maximum Ejected CEA Worth (%ip)

Full Power with Bank 4 inserted worst CEA ejected 0.22 0.30 0.36 0.36 Zero Power with Banks 4+3 inserted worst CEA ejected 0.28 0.48 0.69 0.69 Page 26 of 62

1 l

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

1 2 0.2751 0.2347 3 4 5 6 7 0.3721 0.9116 0.8667 1.0990 0.8673 8 9 10 11 12 13 0.2018 1.0680 1.3302 1.3466 1.0139 1.4143 14 15 16 17 18 19 0.2459 1.1808 1.4544 1.2804 1.3796 1.2332 1.6285 20 21 22 23 24 25 0.7589 1.2985 1.2603 1.3288 1.0904 1.3210 26 27 26 29 30 31 32 1.1522 1.0163 1.3734 1.0589 1.4306 1.2305 33 34 35 36 37 38 39 C-0.9643 1.4475 1.2481 1.3249 1.2349 1.1051 Maximum 1-Pin Peak at 23% Core Height Cycle 14 Assembly Power Distribution Omaha Public Power District Figure 0 MWD /MTU, HFP, Equilibrium Xenon Fort Calhoun Station Unit No.1 5-1 Page 27 of 62

AA - Assembly Location B.BBBB - Assembly Relative Power Density b,e, 1

C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.2661 0.2308 0.3440 0.8433 0.8142 1.0429 0.8104 8 9 10 11 12 13 0.2007 0.9873 1.2064 1.3964 0.9951 1.4327 14 15 16 17 18 19 0.3106 1.1085 1.5300 1.2296 1.5334 1.2628 20 21 22 23 24 25 26 O.3123 27 28 29 30 31 32 1.0861 0.9987 1.5319 1.0716 1.5193 1.2054 33 1.7370 0.3670 34 35 36 37 38 39 0.8815 1.4567 1.2780 1.5311 1.2095 1.0309 Maximum 1-Pin Peak at 23% Core Height Cycle 14 Assembly Power Distribution Omaha Public Power District Figure 7,000 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-2 Page 28 of 62

AA - Assembly Location B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.3175 0.2775 3 4 5 6 7 0.3963 0.9056 0.8645 1.1034 0.8602 8 9 10 11 12 13 0.2426 1.0402 1.1908 1.3758 0.9966 1.3986 14 15 16 17 18 19 0.4078 1.1243 1.4778 1.1698 1,4593 1.1899 1.6550 20 21 22 23 24 25 26 27 28 29 30 31 32 1.1709 1.0130 1.4674 1.0194 1.4047 1.1143 gg 0.4494 34 35 36 37 38 39 0.9477 1.4310 1.2082 1.4289 1.1175 0.9687 Maximum 1-Pin Peak at 17% Core Height Cycle 14 Assembly Power Distribution Omaha Public Power District Figure 14,000 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-3 Page 29 of 62

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

1 1 2 0.2949 0.2571 3 4 5 6 7 0.2572 0.8059 0.8772 1.1819 0.9490 B 9)gy 10 11 12 13 0.1284 103386: 1.1595 1.3600 1.0870 1.5387 Nsd$Y 14 15 16 17 18 19 0'.2191 1.0341 1.3964 1.3206 .1.4719 1.3286 20 21 22 23 24 25 26 O.3633 27 28 -

29 30 31 32

1. '28 1.1100 1.4798 1.1114 '1.4117- 1.1376 3

0.4444 34 35 36 37 38. 39gyj!

1.0791 1.6002 1.3582 1.3824 1.1438 $62655d' 1.7810 MdpyM Maximum 1-Pin Peak at 20% Core Height jf$ - Bank 4 Locations

. Cycle 14 Assembly RPD Bank 4 in Omaha Public Power District Figurc 0 MWD /MTU, HFP, Equilibrium Xenon Fort Calhoun Station Unit No.1 5-4

-Page 30 of 62

. _ , - . _ _ . . _ . _ ~ _ _ _ . - _ . _ . . _ _ . . _ _ . _ _ - . _ _ . _ . . . _ . . _ . _ . _ _ _ . _ _ . _ . . _

e AA - Assembly Location .

B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.3404 0.3033 3 4 5~ 6 7' O.2691 0.7999 -0.8769 0.1847 0.9387 8 9geg; 10 11 12 13 0.1557 40.50481 1.0364 1.3930 1.0665 1.5173 MMP 14 15 16 17 18- 19 ,

0.3647 0.9855 1.4196 1.2066 1.5563 1.2808 20 21 22 23 24 25-26 0.4241 27 28 29 30 31- 32 1.2807 1.0974 1.5749 1.0737 1.0434 1.0474 33 1.8000 34 35 36- 37 38 39 "A 1.0501 1.5686 1.3088 1.4985 1.0519 h)i' 7

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Maximum 1-Pin Peak at 17% Core Height 7---

g - Bank 4 Locations

. Cycle 14 Assembly RPD Bank 4 in Omaha Public Power District Figure 14,000 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-5 Page 31 of 62

l 6.0 THERMAL-HYDRAUUC DESIGN 6.1 QNBR ANALYSIS Steady state DNBR analyses of Cycle 14 at the rated power of 1500 MWt have i been performed using the TORC computer code described in Reference 1 and the CE-1 critical heat flux correlation described in Reference 2. The CETOP-D computer code described in Reference 3 was used in the setpoint analysis, but '

was replaced by the TORC code for DNBR analyses. This is different from the combination that was used in the Cycle 8 through Cycle 13 Fort Calhoun reload ,

analyses (References 4 through 9) in that the more accurate TORC code was used in place of the CETOP-D code. The reload methodology for Cycle 14 can be found in Reference 10.

Table 6-1 contains a list of pertinent thermal-hydraulic parameters used in both safety analyses and for generating reactor protective system setpoint information. The calculational factors (engineering heat flux f actor, engineering factor on het 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 / probability level (Reference 11) 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 f,4WD/MTU bumup is 0.5% in MDNBR. This penalty was applied in the derivation of the SCU MDNBR design limit of _1.18 (References 6 and 12) in the statistical combination of uncertaintles_(Reference 11). The Westinghouse fuel does not have any DNBR penalty associated with the design requirements for the Westinghouse fuel based on NRC fuel bowing requirements, thus, the more limiting CE fuel bow penalty was used in the analyses.

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TABLE 6-1 FORT CALHCUN UNIT NO.1, CYCLE 14 THERMAL HYDRAULIC PARAMETERS AT FULL POWER Unit Cycle 14*

Total Heat Output (Core Only) MWt 1500 10' BTU /hr 5119 L

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 Inlet Temperature F 545 Total Reactor Coolant Flow gpm 202,500 (Steady State) 10' lbm/hr 76.32 (Thrcugh the Core) 10* lbminr 73.06 Hydraulic Diameter (Nominal Channel) ft .044

- Average Mass Velocity 10' Ibm /hr-f t' 2.226 Core Average Heat Flux (Accounts for Heat Generated BTU /hr-f t' 181281 in Fuel Rod)

Total Heat Transfer Surface Area ft' 28,241 "

Average Core Enthalpy Rise BTU /lbm 72.6 Average Unear Heat Rate kW/ft 6.01 *

• Engineering Heat Flux Factor 1.03 * *

• Engineering Factor on Hot Channel Heat input 1.03* *

• 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 14 specific value of 424 fuel displacing shims.

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

Page 33 of 62

~ _ ~ . - -. -. .-. .

7.0 TRANSIENT ANALYSQ This section preseres the results of tne Omaha Public Power District Fort Calhoun Station. Unit 1, Cyc;v 14 Non-LOCA safety analyses at 1500 MWt.

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 ,

Reactor Protection System (RPS) is necessary to prevent exceeding acceptable I rnits.

2. AOOs for which the initial steady state thermal margin, maintainec by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding acceptable limits.

3. Postulated Accidents.

Core parameters input to the safety analyses for evaluating approaches to DNB and centerline temperature to melt fuel design limits are presented in Table 7-2.

As indicated in Table 7-1, no reanalysis was performed for the DBEs for which key transient input parameters are within the bounds (i.e., conservative with respect to) of the reference cycle values (Fort Calhoun Updated Safety Analysis Report including Cycle 13 analyses, Reference 1). For these DBEs the results and conclusions quoted in the reference cycle analysis remain valid for Cycle 14.

For those analyses indicated as reviewed, calculations were performed in accordance with Reference 6 until a 10 CFR 50.59 determination could be made that Cycle 14 results .

would be bounded by Cycle 13 or the USAR reference cycle.

Events v>ere 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 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 presentations of the results of the reanalyses are provided in Sections 7.1 through 7.3.

Page 34 of 62

- . . . = - . - .. -

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- TABLE 7 FORT CALHOUN UNIT NO.1, CYCLE 14 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary to prevent exceeding acceptable limits:

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

I 7.1.3 Loss of Feedwater Flow Not Reanalyzed 5 7.1.4 Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed 5 i 7.1.5 Startup of an inactive Reactor Coolant Pump Not Reanalyzedi l 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 Excess Load Reanalyzed 2 7.2.2 Sequential CEA Group Withdrawal Reanalyzed 2 7.2.3 Loss of Coolant Flow Reviewed 3.5 7.2.4 CEA Drop Reanalyzed 7.2.5 Boron Dilution Reviewed 7.2.6 Transients Resulting from the Malfunction of One Steam Generator Not Roanalyzed4 7.3 Postulated Accidents 7.3.1 CEA Ejection Reanalyzed 7.3.2 Steam Line Break Reviewed 5 7.3.3 Seized Rotor Reanalyzed 5

-7.3.4 Steam Generator Tube Rupture Not Reanalyzed 1

Technical Specifications preclude this event during operation.

2 Requires High Power and Variable High Power Trip.

3 Requires Low Flow Trip.

4 Requires 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.

Page 35 of 62

TABLE 7-2 FORT CALHOUN UNIT NO.1, CYCLE 14 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Physics Parameters Units Cycle 13 Values Cycle 14 Values Radial Peaking Factors For DNB Margin Analyses (F;)

Unrodded Region 1.70* 1.78*

Bank 41nseded 1.73* 1.91*

For Planar Radial Component (FL ) of 3-D Peak (CTM Limit Analyses)

Unrodded Region 1.75* 1.85*

Bank 4 Inserted 1.77* 2.0*

Maximum Augmentation Factor 1.000 1.000 Moderator Temperature Coefficient 10-4 Ap/ F -2.7 to +0.5 -3.0 to +0.5 Shutdown Margin (Value Assumed in Limiting EOC Zero Power SLB) %Ap - 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 36 of 62

TABLE 7-2 (Continued)

Safety Parameters Units Cycle 13 Values Cycle 14 Values Power Level MWt 1500* 1500*

Maximum Steady State Temperature 'F 543* 545*

Minimum Steady State Pressurizer Pressure psla 2075* 2075*

Maximum Augmentation Factor 1.000 1.000-Reactor Coolant Flow gpm 202,500* 202,500*

Steam Generator Tube Plugging  % 6 6 Negative Axial Shape LCO Extreme Assumed at Full Power (Ex-Cores) asiu -0.18 -0.18 Maximum CEA insertion  % insertion at Full Power of Bar:i< 4 25 25

. Maximum Initial Linear Heat Rate for Transient

' Other than LOCA kW/ft 14.4 13.8 Steady State Linear Heat Rate for Fuel CTM Assumed in the Safety Analysis kW/ft 22.0 22.0 CEA Drop Time to 100%

including Holding Coil Delay sec 3.1 3.1 Minimum DNBR (CE-1) 1.18* 1.18*

The effects of uncertainties on these parameters were accounted for statistically in the DNOR and CTM calculations. The DNBR analysis 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 37 of 62

TABLE 7-3 FORT CALHOUN UNIT NO.1 DESIGN BASIS EVENTS REANALYZED FOR CYCLE 14 Reason foi Acceptance Summary Event Reanalysis Criteria of Results Sequential CEA Calculate cycle specific Minimum DNBR 2 MDNBR =1.72 Group Withdrawal ROPM values 1.18 using the CE-1 PLHGR< 22 kW/ft correlation. Transient PLHGR $ 22 kW/ft.

CEA Drop incorporated bounding Minimum DNOR 2 MDNBR = 1.38 input values 1,18 using CE-1 PLHGR < 22 kW/ft correlation. Transient.

PLHGR $ 22 kW/ft Excess Load Reclassified as a ROPM event Minimum DNBR ;> MDNBR = 1.31 (methodology change) 1.18 using CE-1 PLHGR < 22 kW/ft correlat:on. Transient PLHGR $ 22 kW/ft RCS Depressurization -To provide a conservative Pbias Pbias value $ the Pbias = 30 pstr.

Input for the TM/LP due to the previous cycle's limiting Excess Load methodology value (from Excess Load change and RCS Depressurization)

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7.0 TRANSIENT ANALYSIS (Continued) 7.1 ANTICIPATED OPERATIONAL OCCURPENCES (CATEGORY 1) 7.1.1 RCS Depressurization Event The RCS Depressurization event was reanalyzed for Cycle 14 to ,

determine the pressure bias term for the TM/LP trip setpoint.

The RCS Depressurization event is one of the Design Basis Events analyzed to determine the maximum pressure bias term input to the TM/LP trip. The methodology used for Cycle 14 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 temperature and RCS pressure during the transient are monitored by the TM/LP trip directly.

Consequently, with TM/LP trip setpoints and the bias term determined in this analysis, adequate protection will be provided for the RCS Depressurization event to prevent the acceptable DNBR design limit from being exceeded. Table 7.1.1-1 provides a sequence of events for the RCS Depressurization analysis.

The analysis of this event shows that incorporating a pressure bias term of 30 psia in the TM/LP trip setpoints will ensure that the RPS provides adequate protection to prevent the acceptable DNBR design limit from being exceeded during an RCS Depressurization event.

The RCS Depressurization event is the only event that is currently analyzed to determine the pressure bias term, since the Excess Load event has been reclassified as an event requiring initial margin for protection. The Excess Load event is discussed in section 7.2.1.

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. . . - - - - - .. . . - -. .-. . = - - - .

I TABLE 7.1.1 - 1 FORT CALHOUN UNIT NO.1; CYCLE 14 SEQUENCE OF EVENTS FOR RCS DEPRESSURIZATION Time (sec) Event Setpoint or Value 0,000 Inadvertent Opening of Both ------

Pressurizer Power Operated Relief Valves 7.382 Reactor Trip 2075.75 psia 9.409 Time of Minimum DNBR 2047.16 psh Page 40 of 62

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) 7.2.1 Excess Load Event The Excess Load event was reclassified for Cycle 14 from an event which is protected by an RPS trip to an event which is protected by the RPS and sufficient initial thermal margin which is maintained by the LCOs. This reclassification does not result in a net gain in margin. It only transfers the margin requirements from the LSSS to the LCO.

The Excess Load event was analyzed for Cycle 14 to determine the DNB and LHR ROPMs which are used to ensure sufficient margin is included in the DNB and LHR LCOs to provide protection to the fuel design limits in the event of an Excess Load event. The methodology used to perform the analycis is described in Reference 6. The key input parameters used in the Cycle 14 Excess Load analysis are presented in Table 7.2.1 -1.

It is assumed in the analysis that the reactor will trip on Variable High Power during an excess load event. Therefore, the key to the analysis is maximizing the time between the initiation of the event (instantaneous opening of the steam dump and bypass valves) and the time at which the Variable - High Power trip (VHPT) _ signal is generated. Several assumptions are made to maximize this time. Since the VHPT uses the auctioneered higher value of the excore power signal and AT-Power calculator, an MTC is chosen which ensures that the AT-Power calculator and the excore detectors both reach the VHPT setpoint at the same time.

The maximum temperature _ shadowing factor is used to maximize the decalibrationof theexcoredetectorsdueto RCScooldown. Also,thetime constants for the but and cold leg resistance temperature detectors (RTDs) are chosen to maximize the lag between the AT-Power calculator and the actual core heat flux.

The DNB and LHR ROPMs calculated for the Excess Load event are

- compared to those calculated for other AOO events such as the CEA Drop and CEA Withdrawal to determine the most conservative (largest) ROPMs to input to the calculatio_n of the LCOs. This ensures that there will be sufficient margin included in the LCOs to protect all AOO events requiring initial margin for protection, it was concluded from the Cycle 14 analysis that the ROPM required by _

the Excess Load event was bounded by the requirements of the CEA Drop Event, Page 41 of 62

TABLE 7.2.1 - 1 FORT CALHOUN UNIT NO.1, CYCLE 14 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD ANALYSIS Parameter Units Cycle 14 Initial Core Power Level MWt 1530 Core inlet Coolant Temperature 'F 547 Pressurizer Pressure psia 2053 Moderator Temperature Coefficient x 10-4 Ap/* F -0.707 Doppler Coefficient Multiplier 0.85 CEA Worth at Trip %Ap 5.7922 Excore Temperature Shadowing Factor  %/* F 0.35 Cold Leg RTD Time Constant see 12.0 Hot Leg RTD Time Constant sec 3.0 i

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7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) 7.2.2 CEA Withdrawal Event The CEA Withdrawal (CEAW) ovent was reanalytM for Cycle 14 to determine the initial margins that must be maintained by the Limiting Conditions for Operations (LCOs) such that the DNBR and fuel centerline to melt (CTM) design limits will not be exceeded in conjunction with the RPS (Variable High Power, High Pressurizer Pressure, or Axial 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 sudicient initial steady state thermal margin provided by the DNBR and Linear Heat Rate (LHR) related LCOs.

Fur the HFP CEAW DNBR analysis, a MTC value identical to that utilized in Reference 8 and a gap thermal conductivity consistent with the assumption of Reference 6 were used in conjunction with a variable reactivity insertion rate.

The HFP case for Cyc!e 14 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 DNB and PLHGR LCOs. Since a negative 10 CFR 50.59 determination was made for Cycle 14, the conclusions for Cycle 12 remain valid and applicable to Cycle 14.

The zero power case was analyzed to demonstrate 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 are 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 5.46 which is less than the Cycle 12 value of 6.99, but still farin 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 conterline melt limit. The key input parameters used for the zero power case are presented in Table 7.2.2-1.

Page 43 of 62

7.0 -TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.2 CEA Withdrawal Event (

• ainued)

It may be concluded that the CEA Withdrawal event, when initiated from the Technical Specification LCOs (in conjunction with the Variable High Power Trip, if required), will not lead to a DNBR or fuel temperature which violates the DNBR and CTM design limits. It was further concluded that the initial available overpower margin requirements for this event were bounded by that of the CEA Drop event.

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TABLE 7.2.2- 1 FORT CALHOUN UNIT NO.1, CYCLE 14 KEY PARAMETERS ASSUMED IN THE HZP CEA WITHDRAWAL ANALYSIS Parameter Units Cycle 12 Cycle 14 Initial Core Power Level MWt 1 1*

Core inlet Coolant Temperature *F 532 532*

Pressurizer Pressure psia 2053 2075*

Moderator Temperature Coefficient x 10-4 Ap/ F + 0.5 + 0.5 Doppler Coefficient Multiplier 0.85 0.85 CEA Worth at Trip %Ap 5.28 6.407 Reactivity insertion Rate Range x 10-4 Ap/sec 0 to 1.0 0 to 2.7 CEA Group Withdrawal Rate in/ min 46 46 Holding Coil Delay Time sec 0.5 0.5 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.

Page 45 of 62 .

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.3 Loss of Coolant Flow Event The Loss of Coolant flow event was reviewed for Cycle 14 and it was determined that the event was bounded by the reference cycle (Cycle 12) analysis. The input parameters are listed for Cycles 12 and 14 for comparison in Table 7.2.3-1.

'Thus, it was concluded that the reference cycle analysis is bounding for Cycle 14 operation.

Page 46 of 62

TABLE 7.2.3- 1 FORT CAUiOUN UNIT NO.1, CYCLE 14 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Param I Units Cycle 12 Cycle 14 Initial Core Power Level MWt 1500* 1500*

Initial Coro itslet Coolant Temperature 'F 545* 545*

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

Pressurizer Pressure psia 2075* 2075*

Moderator Temperature Ccefficient x 10-4 Ap/'F + 0.5 +0.5 Doppler Temperature Multiplier 0.85 C.85 CEA harth at Trip (ARO) %Ap -0.50 -0.72 LFT Analysie Setpoint  % of initial flow 93 93 LFT Response Time sec 0.05 0.65 CEA Holding Coil Dalay sec 0.5 0.5 CEA Time to 100% Insertion sec 3.1 3.1 (Including Holding Coil Delay Total Unrodded Radial Peaking Factor (F;) 1.80 1.78 Tht uncertainties on these parameters were combined statistically rather than deterministically. The values listed represent the bounds included in the statistical combination.

Pago 47 of 62

7.0 TRANSIENT ANALYSIS (Continued) [

7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.4 Full Length CEA Drop Event The Full Longth CEA Drop event was rcanalyzed for Cyclo 14 to detc rmino the initial margins that must be maintained by the Limning Conditions for Operations (LCOs) such that the DNBR and fuel CTM design limits will not be excooded.

L This event was analyzed parametrically in initial axial shape and rod configuration using the methods described in Referenco 6. Tablo 7.2.4- 1 lists tho lf cy input paramotors used for Cycle 14 and compares thom to the reference cyclo (Cycle 11) values whilo Tablo 7.2.4-2 contains a sequenco of events for the CEA Drop analysis.

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

-0.182, which is outsido of the LCO limit of -0.08. This resulte in a minimum CE-1 DNBR of 1.377. A maximum allowablo initiallinear heat generation rato of 18.4 kW/ft could exist as an Inlllal cc,..dition without excooding the acceptable fuel CT M limit ef 22 kW/ft during this transient.

This amount of margin is assured by setting the LHR related LCOs based on the more limiting allowablo LOCA linear heat rato. -

It can be concluded that the CEA Drop event was the most limiting of tho .

AOO's dependent upon initial availablo overpower margin. When initiat 'd from the Technical Specification LCOs, the event will not excood the DNBR CTM design limits.

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TABLE 7.2.4- 1 FORT CALHOUN UNIT NO.1. CYCLE 14 KEY PARAMETERS ASSUMED IN THE HFP CEA DROP ANALYSIS Parameter Units Cyclo 11 Cyclo 14  ;

Initial Coro Power Lovel MWt 1500* 1500* t Coro inlet Coolant Temperature 'F 543* 545*

Pressurizor Pressure psia 2075* ?075*

- Coro Mass Flow Rato gpm 202,500* 190,000*

Moderator Temperaturo Coefficient x 10-4 Ap/'F - 2.7 - 3.0 Doppler Coefficient  ;

Mult plier 1.15 1.40 CEA Insertion at Maximum Allowed Power  % Insertion of Bank 4 25 25 Dropped CEA Worth Unrodded, %Ap - 0.2337 - 0.2947 PDIL, %Ap - 0.2205 - 0.2040 Maximum Allowed Power '

Shapo Index at Negativo Extremo of LCO Band -0.18 -0.18 Radial Peaking Distortion Factor Unrodded Region 1.1566 1.1937 Bank 4 Inseded 1.1598 1.1004 The DNBR calculations used the methods discussed in Section 0.1 of this document and ,

detailed in References 2 through 5. The ollects of uncertainties on these parameters

were accounted for statistically in the DNBR and CTM calculations.

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TABLE 7.2.4-2 FORT CAUiOUN UNIT NO.1, CYCLE 14 SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP Time (sec) Event Setpoint or Value 0.0 CEA Begins to Drop into Core ---

1.0 CEA Reaches Fully inserted Position 100% insertion 1.14 Core Power Level Reaches a Minimum G3.8% of 1500 MWt and Begins to Return to Power Due to Reactivity Feedbacks 78.97 Core inlet Temperature Reaches a 538.68'F Minimum Value 199.9 RCS Pressure Reaches a Minimum 1996.10 psia VaJe 200.0 Core Power Returns to its Maximum 94.85% of 1500 MWt Value 200.0 Minimum DNBR is Reached 1.377 (CE-1 Correlation)

Page 50 of 62

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.5 Boron Dilution Event The Boron Dilution event was reviewed for Cycle 14 to verify that sufficient timo is available for an operator to identify the causo and to terminato a boron dilution ovent for any modo of operation before SAFDL limits aro violated.

Table 7.2.5-1 compares the values of the key transient parnmeters assumed in each modo of operation for Cycle 14 and the reference cycle.

Cycle 13. The Cycle 14 analysis utilized a mass basis in the calculations, as was used in Cyclo 13, rather than a volumetric basis to ensure that all operating temperaturo ranges for all modes of operation were bounded.

Since the critical boron concentration for Cycle 14 is less than the corresponding Cycle 13 values for all modes there is no further analysis required as the Cycle 13 results will bound Cyclo 14.

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_ _ _ _ . _ . _ _ . . ~ _ _ . _ _ - _ _ _ . _ . _ . . _ . _ . _ . _ . _ _ _ _ _ . _ . . _ . _ .

TABLE 7.2.5- 1  !

FORT CAUiOUtl UNIT NO.1, CYCLE 14 '

KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS 1

Parameters Cycle 13 Values Cycle 14 Values Critical Boron Concentration, ppm _ (ARQ, No Xenon)

Mode  ;

Hot Standby 1662 1292 Hot Shutdown 1662 1292 Cold Shutdown - Normal RCS Voluren 1457 1204 Cold Shutdown - Minimum RCS Volumc 1279 1204 Refueling 1454 1180 Inverso Boron Worth, ppm /%Ap l

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 Margin Assumed, %Ap Mode L

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)** 1900 1900 Shutdown Groups A and B out, all Regulating Groups inserted except most reactive rod stuck out.

Includes a 5.0%Ap shutdown margin.

l Page 52 of 62

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7.0 TRANSIENT ANALYSIS (Continued) f 7.3 POSTULATED ACCIDENTS l 7.3.1 CEA Ejection '

The CEA Ejection event was reanalyzed for Cyclo 14 since Westinghouse  !

will bo providing a now fuel design. A summary report was transmitted to  :

the NRC for review in Reference 14.  ;

l 7.3.2 Steam Lino Break Accident This accident was reviewod for Cycle 14 using the methodology t discussed in References 6 and 12. The Stcam Uno Broak (SLB) accident was previously analyzed in the Fort Calhoun FSAR and satisfactory rosults  :

- were reported therein. The SLB accidents at both HZP and HFP woro '

examined in the referenco cyclo (Cyclo 8) saiety evaluation with acceptable results obtained. Both the FSAR and referenco cyclo [

ovaluations are reported in the 1991 updato of the Fort Calhoun Station f Unit No.1 USAR. ,

The Full Power Stcam Lino Break accident was reviewed for Cycle 14 for a more negative MTC of -3,0 x 10-4 Ap/*F than the -2.5 x 10-4 Ap/* F i value that was used in the Cyclo 8 analysis. However, the cooldown curve -

for Cycle 14 is bounded by Cyclo 8 (as shown in Figure 7.3.2-1). This figure shows that the reactivity insertion for the Cycle 14 core with an MTC of -3.0 x 10-4 Ap/*F due to a SLB accident at full power is substantially less than the value used in the Cycle 8 analysis. (This smaller reactivity insertion is due to the uso of the DIT cross-sections which are valid for a range of moderator temperatures from room temperaturo to 600'K whito the analyses prior to Cycle 9 were performed with cooldown curves- ,

derived by conservatively extrapolating CEPAK cross-soction values to '

low temperatures.) The Cycle 14 minimum available shutdown worth at '!

HFP is 6.2885 %Ap compared to a Cycle 8 value of 6.68%Ap. This implies

'a margin decrease of 0.395%Ap. The Cycle 14 moderator cooldown reactivity botween 574* F and 350* F at HFP is 4.7%Ap comparod to 5.37 .

%Ap in Cycle 8.- This implies a margin inerease of 0.67 %Ap. The Cycle 14 dopoler coefficient is more negative than the Cycle 8 doppler including uncerdnties. However, this loss in margin is offset by the gain in margin - i

- from the n;oderator cooldown reactivity. The not gain ensuros that tho ,

overall reactivity insertion for a Cycle 14 SLB is less than that of the roferonce cycle analysis, Theroforo, the return to power is less than that of the referenco cycle and Cycle 1 FSAR analysos.-

F L

l Page 53 of 62 h

L. .,...,-,_-_____- - ,_. _ __,_ ,. _ ,..,_,_._ _ _ ___ _ ,._ _ _-_ _ ._ ,_.__ . _ , _ _ , , , _ , . _ _ . . _ . _ , _ , , _ _

l 7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) 7.3.2 Steam Uno Break Accident (Continued)

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

Again, the Cyclo 14 cooldown for an MTC of -3.0 x 10-4 Ap/'F shows a ,

substantially smaller reactivity insertion than was used in the Cycle 8  :

analysis (as seen in Figure 7.3.2-1). Sinco the minimum availablo shutdown margin for Cycle 14 remains unchanged from the referenco  !

cycle vr'w (4.0%Ap), the overall reactivity insertion for the Cyclo 14 SLB accidentwill bo less severo than that reported for the referenco cyclo and the FSAR (Cyclo 1) casos.

Based on the ovaluation presented above, it is concluded that the consequences of a SLB accident initiated at either 7ero or full power are less sovero than the referenco cycle and FSAR (Cycle 1) cases.

Since a negativo determination utilizing the 10 CFR 50.59 criteria was i

made for the Cycle 14 SLB accident, no reanalysis was performed. Thus, it was concluded that the teforence cycle analysis is bounding ior Cycle 14 operation.

Page 54 of 62

f 1

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1. ~

l- N - d>

N R Nx N NYaum  :

ca i4 mP/

i

' mm

~

b.m neo.m em 4m.m 4m.m sm.m amm ez =

CORE AVERAGE MODERATOR TEMPERATURE (F) l l Steam Line Break Accident- Omaha Publi': Power District Figuri Reactivity vs. Moderator Temperature Fort Calhoun Station Unit No.1 7.3.2 - 1 Page 55 of 62 N N

7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) ,

7.3.3 Solzod Rotor Event .

The Solzod Rotor event was reanalyzed for Cyclo 14 to demonstrate that  :

only a small fraction of fuel pins aro predicted to fail during this event. The analysis showed that Cycle 141s bounded by tho roforenco cyclo (Cyclo 9)  :

analysis becauso an Fa of 1.85 was assumed in the Cyclo 9 analysis and the Cycle 14 Technical Specification of 1.78 remains conservative with respect to the F1 value used in the Cyclo 9 analysis.

Thoroforo, the total number of pins prodlcted to fall will continue to be less than 1%of allof thefuelpinsinthecoro. Basedonthisrosult,therosultant ,

site boundary doso would be well within the limits of 10 CFR 100.

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Page 56 of 62

t 8.0 ECCS PERFORMANCE ANALYSIS Both Cycle 14 Largo and Small Break Loss of Coolant accident analyses woro

• performed by Westinghouse using the methodology discussed in Referenco 1. A summarycontainingthotosultsof theanalysoswassubmittedin Referenco2. Thepeak linear heat generation rato of 15.5 kW/It was conservatively roduced to 13.8 kW/ft for the non- LOCA transients to onsuro tho C E fu el mechanical design roq uirements wor o valid for the operation of Cycle 14.

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9.0 STARTUP TESTING l I

The startup testing program proposed for Cyclo 14 is identical to that used in Cycle 13. It is also the same as the program outlined in the Cycle 6 Reload Application, with two exceptions. First, a CEA exchango techniquo (Reference 1) for zero power rod worth measurements will be performed in accordance with Referenco 2, replacing the boration/ dilution method. Also, low power CECOR flux maps and pseudo-ejection tod measurements will be substituted for the full coro symmetry checks.

The CEA exchange techniquo is a method for measuring rod worths which n. both faster and produces loss wasto than the typical boration/ dilution method. The startup testing method used in Cycles 11,12 and 13 employed the CEA exchange technique  ;

exclusively. Results from the CEA exchango technique wero within the acceptanco and  ;

review critoria for low power physics parameters. The combination of the pseudo-ejection techniquo at zero power and low power CECOR maps provides for a less time censuming but equally valid technique for detecting azimuthal power tilts during roioad core physics testing. The pseudo-ejection rod measurernent involves the dilution of the lead bank (Bank 4) into the core, borating a Bank 4 CEA out, and then exchanging (rod swap) the CEA against other symmetric CEA's within Bank 4 to measure rod worths.

The acceptance and review critoria for those tests are:

Test Acceptanco Critoria Review Cr!teria CEA Group Worths 15% of prodleted 15% of predicted Pseudo-eJoction None 1.5e deviation from rod worth group averago measurement Low Power CECOR Technical Specifica- Azimuthal tillless than maps tion limits of F;, 20%.

FL , and T, i OPPD has reviewed ihese tests and has concluded that no unreviewed safety question l exists for implementation of those procedures, i

Page 58 of 62

10.0 REFERENCES

References (Chapters 1-5)

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

OPPD-NA-8301-P, Revision 04, April 1991.

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

3. " Omaha Public Power District Roload Coro Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-P, Revision 03, March 1991. ,

4. " Omaha Batch M Reload Fuel Design Report", CEN-347(O)-P Revision 1 -R January 1987.

5. " Westinghouse Roload Fuel Mechanical Design Evaluation for the Fort Calhoun Station Unit 1", WCAP-12977 (Propriotary), June 1991.

Page 59 of 62

i 10.0 REFEhENCES (Continued)

References (Chapter 6)

1. " TORC Codo, A Computer Codo for Determining the Thermal Margin of a Reactor Coro". CENPD-101 -P, July 1975.

2.

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

3. "CETOP-D Codo Structure and Modeling Methods for Calvert Clifis Units 1 and 2", CEN-191-(B)-P December 1981

4. Safety Evaluation by the Offico of Nuclear Reactor Regulation Supporting Amendment No. 70 to Facility Operating Uconso 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 Offico of Nuclear Reactor Regulation Supporting Amendment No. 77 to Facility Operating Uconso 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 Amendment No. 92 to Facility Operating Uconso 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 Amendment No.109 to Facility Operating Uconso 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 Regulation Supporting Amendment No.117 to Facility Operating Ucense No. DPR-40 for Omaha Public Power District, Fort Calhoun Station, Unit No.1, Docket No. 50-285, December 14,1988.

9. Safety Evaluation by the Offico of Nuclear Reactor Regulation Supporting Amendment No.126 to Facility Operating Ucense No. DPR-40 for Omaha Public Power District, Fort Calhoun Station, Unit No.1, Docket No. 50-285, April 4,1990.

10. " Omaha Public Power District Reload Coro Analysis Methodolcijy Overview",

OPPD-NA-8301-P, Revision 04, March 1991,

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

12. 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.

Page 60 of 62

10,0 REFERENCES (Continued)

References (Chapter 7)

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

2. " Statistical Combination of Uncertainties Methodology, Part 1: Axlal Power Distribution 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, Par 12: Combination of System Paramet9r Uncertaintes in Thermal Margin Analysis for Fort Calhoun Unit 1", CEN-257(0)-P, November 1983.

4. " Statistical Combination of Uncertainties Methodology, Part 3: Departuro from Nucleate Bolling and unear Heat Rate Limiting Conditions for Operation for Fort Calhoun", 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 Reload Coro Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-P, Revision 03, March 1991.

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

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

9. "CESEC, DigitalSimulation of a Combustion Engineering Nuclear Steam Supply System", Enclosure 1-P to LD-82-001, January 6,1982.

10. " Response to Questions on CESEC", Loulslana Power and Ught Company, Waterford Unit 3, Docket 50-382, CEN-234(C)-R December 1982.

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

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

13 Letter UC-89-1172, K. J. Morris (OPPD) to Document Control Desk (NRC),

dated November 8,1989.

14 Letter UC-91-198R, W. G. Gates (OPPD) to Document Control Desk (NRC),

dated July 31,1991.

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