ML20046A083

Omaha Public Power District Fort Calhoun Station Unit 1 Cycle 15 Reload Evaluation.

ML20046A083
Person / Time
Site:	Fort Calhoun
Issue date:	07/19/1993
From:	OMAHA PUBLIC POWER DISTRICT
To:
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ML20046A082	List: ML20046A081 ML20046A083
References
NUDOCS 9307260162
Download: ML20046A083 (72)

Text

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Omaha Public Power District. i Fort Calhoun Station Unit No.1 l 1

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l Cycle 151 Reload Evaluation I

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x FORT CALHOUN STATION UNIT NO.1 ,

CYCLE 15 14ELOAD EVALUATION TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND

SUMMARY

... . . . .... .. ..4 2.0 OPERATING HISTORY OF CYCLE 14 .. .. ... . ..... . .. .. .5 3.0 GENERAL DESCRIPTION . .... . .... . .. . . . ..6 4.0 FUEL SYSTEM DESIGN . . . . .. . ... ... ..... .. 21 5.0 NUCLEAR DESIGN .. .. .... ....... . ......... . .. .. . 22 -

5.1 PHYSICAL CHARACTERISTICS . . .. .. . ... .. .... . . .. 22 5.1.1 Fuel Management . . . . . . . . . . .... .. ........ 22 5.1.2 Power Distribution . . . .. . . .. . .. .. ..

. 23 5.1.3 Safety Related Data .... . . ....... .. . ... .. . 24 5.1.3.1 Ejected CEA Data . . . . . . . . . . . . ...... .. . . . . 24' 5.1.3.2 Dropped CEA Data .. . . . ...... ... . . .. . . ...... 24 5.2 ANALYTICAL INPUT TO INCORE MEASUREMENTS .... . . . .. 24 5.3 NUCLEAR DESIGN METHODOLOGY . . . . . . . . . . .......... .... 24 5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS . . . ... 24 6.0 THERMAL-HYDRAULIC DESIGN . . ..... .. .. . ..... . ...... 33 6.1 DN BR ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. ..... 33 6.2 FUEL ROD BOWING . . . . . . . . ........... . .. .............. 33 Page 2 of 67L

FORT CALHOUN STATION UNIT NO.1 CYCLE 15 RELOAD EVALUATION TABLE OF CONTENTS (Continued)

Page 7.0 TRANSIENT ANALYSIS . .. . . ... . .. ... ... . 35 7.1 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 1) . . 40 7.1.1 RCS Depressurization Event ...... . . . .. ..... 40 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) . 42 7.2.1 Excess Load Event . .. .. .... .. . ... .. 42 7.2.2 CEA Withdrawal Event . . .... ..... .. . .... . ... 45 7.2.3 Loss of Coolant Flow Event . . . ....... .. .. ...... 50 7.2.4 Full Length CEA Drop Event . .. . . . .. . ... ... 52 7.2.5 Boron Dilution Event ... .. . .. .. . .. 55 7.3 POSTULATED ACCIDENTS . . . . . . . . . . . . . . . . . .. .. . . . 58 7.3.1 CEA Ejection .... . .. .. . . .... .. .. . 58 7.3.2 Steam Line Break Accident . ..... ...... ...... . .. . 58

-1 7.3.3 Seized Rotor Event . . . . . . . . ..... .. . ......... .. 61 8.0 ECCS PERFORMANCE ANALYSIS . . . . . . . . . . . . ..... ........... 62 9.0 STARTUP TESTING . . . ....... . . ...... ........ .......... .. . ~63

10.0 REFERENCES

...... ...... . .... . .. .... ....... .... . 64 Page 3 of 67

1.0 INTRODUCTION

AND

SUMMARY

This report provides an evaluation of the design and performance for the operation of )

Fort Calhoun Station Unit No.1 during its fifteenth fuel cycle at a full rated power of 1,500 MWt. Planned operating conditions remain the same as those for Cycle 14, unless otherwise noted.

The core will consist of 92 presently operating Batches N, P and R assemblies and 41 fresh Batch S assemblies.

The Cycle 15 analysis is based on a Cycle 14 termination point of 13,500 MWD /MTU. In performing analyses of design basis events, limiting safety system settings and limiting conditions for operation, key parameters were chosen to assure that expected Cycle 15 conditions would be enveloped. The analysis presented herein will accommodate a Cycle 15 length of up to 13,500 MWD /MTU with a coastdown of an additional 700 MWD /MTU.

The evaluation of the reload core characteristics has been conducted with respect to the Fort Calhoun Station Unit No.1 Cycle 14 safety analysis described in the 1992 update of the USAR, hereafter referred to as the " reference cycle" in this report unless noted otherwise.

Specific core differences have been accounted for in the present analysis. In all cases, it .

has been concluded that either the reference cycle analyses envelope the new conditions or the revised analyses presented herein continue tc show acceptable results. Where dictated by variations from the previous cycle, changes are being incorporated into the Cycle 15 Core Operating Limits Report.

The Cycle 15 core has been designed to minimize the neutron flux to limiting reactor pressure vessel welds to reduce the rate of RTm shift on these welds. This will maximize the time to reach the screening criteria consistent with the procedure for calculating the amount of radiation embrittlement that a reactor vessel receives, as given in Regulatory Guide 1.99, Revision 2 and 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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2.0 OPERATING HISTORY OF CYCLE 14 Fort Calhoun Station is presently operating in its fourteenth fuel cycle utilizing Batches M, N, P and R fuel assemblies. Fort Calhoun Cycle 14 operation began when criticality was achieved on May 3,1992, and full power was reached on May 9,1992. The reactor has operated up to the present time with the core reactivity, power distributions, and peaking factors having closely followed the calculated predictions.

It is estimated that Cycle 14 will be terminated on or about September 18,1993. The Cycle 14 termination point can vary between 12,000 MWD /MTU and 13,500 MWD /MTU and still be within the assumptions of the Cycle 15 analyses. As of May 18,1993, the Cycle 14 core average burnup had reached 9,700 MWD /MTU.

I Page 5 of 67-

3.0 GENERAL DESCRIPTION The Cycle 15 core will consist of the number and type of assemblies and fuel batches shown in Table 3-1. One Batch M assembly,24 Batch N assemblies and 16 Batch P assemblies will be discharged this outage. They will be replaced by 4 fresh Batch S1 assemblies (3.85 w/o average enrichment),4 fresh Batch S2 assemblies (3.85 w/o average enrichment with 28 IFBA rods at 0.003 gm B io/ inch), 6 fresh Batch S3 assemblies (3.85 w/o average enrichment with 48 IFBA rods at 0.003 gm Bi o/ inch),10 fresh Batch S6 assemblies (3.35 w/o average enrichment with 28 IFBA rods at 0.003 gm B io/ inch),9 fresh Batch S7 assemblies (3.35 w/o average enrichment with 48 IFBA rods at 0.003 gm B i o/ inch) and 8 fresh Batch S8 assemblies (3.35 w/o average enrichment with 64 IFBA rods at 0.003 gm B o/ i inch).

Figure 3-1 shows the fuel management pattern to be employed in Cycle 15. The fuel management strategy for Cycle 15 is identical to the strategy used for Cycle 14. The overall fuel management scheme is designed to maximize the reduction in neutron leakage seen by the reactor vessel and limiting vessel weld locations. This strategy is called " extreme low radial leakage fuel management" and is similar to the fuel management previously used in the Cycle 10 core loading pattern. Listed below are the key parameters that comprise the extreme low radial leakage fuel management strategy:

1) Twelve thrice-burned fuel assemblies on the core periphery will contain four-full-length hafnium flux suppression rods per fuel assembly to locally reduce neutron flux near the limiting reactor vessel welds. Each of the hafnium rods will be l

placed in one of the outer CEA guide tubes of peripheral fuel assemblies.

2) Four once-burned natural uranium fuel assemblies will be located on the core periphery adjacent to the reactor vessel limiting welds. These four peripheral assembly locations could not support the use of full-length hafnium. flux suppression rods due to the .esidence of CEA Shutdown Group A rods.

3) Use of an integral fuel burnable absorber (IFBA) instead of the traditional fuel displacing poison rods within selected new fuel assemblies. The IFBA rods consist of fuel pellets treated with an electrostatically applied, zirconium-diboride coating which encompasses the fuel pellet circumferential surface area. By using IFBA rods, extrerne low radial leakage fuel management can provide greater reduction in vessel flux by increasing the number of fuel rods available to produce the rated power of 1,500 MWt. This will gain radial peaking factor margin that is needed to absorb the inward roll of the core power distribution caused, in part, by the peripheral flux reduction.

Page 6 of 67

- . .-_.c- .. . . __ __ . _ -.

3.0 GENERAL DESCRIPTION (Continued)

The fuel rod and poison rod locations in Batch N shimmed assemblies are shown in Figure 3-2. Figure 3 -3 shows the fuel rod 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 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 average initial enrichment sufficient to achieve the Cycle 15 design exposure. Figures 3-5 through 3-10 provide a diagram of each type of Batch R and S assembly.

The average initial enrichment of the 41 fresh Batch S assemblies is 3.52 w/o U-235, a reduction of 0.07 w/o from Cycle 14. For the third consecutive cycle, the fuel assembly zone loading technique is used to lower the radial power peaking factors within Batches S1 through S8. Batch S1 through S3 assemblies have fuel rods at both 4.0 w/o enriched U-235 and 3.5 w/o enriched U-235, while Batch S6 through S8 assemblies have fuel rods at both 3.50 w/o enriched U-235 and 3.00 w/o enriched U-235.

Figure 3-11 shows the beginning of Cycle 15 assembly burnup distribution for a Cycle 14 termination burnup of 13,500 MWD /MTU. The initial enrichment of each fuel assembly is also shown in Figure 3-11. Figure 3-12 shows the projected end of Cycle 15 assembly burnup distribution. The end of Cycle 15 core average exposure, including coastdown, will be approximately 30,015 MWD /MTU.

4 Page 7 of 67  ;

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TABLE 3-1 FORT CALHOUN UNIT NO.1 CYCLE 15 CORE LOADING Initial BOC EOC Poison IFBA Poison Assembly Number of Avg. Burnup* Avg. Burnup** Rods per Rods per Loading Designation Assemblies MWD /WU MWD /MU Assembly Assembly gm B30/in.

N 8 28,551 32,902 0 -

0 ,

N/ 8 36,404 39,035 8 -

0.020 P 8 29,759 45,753 0 -

0 P/ 16 32,732 42,623 8 -

0.027 R1 4 4,222 8,751 -

0 -

R2 16 13,355 28,013 -

28 0.003 R3 4 19,261 38,763 -

48 0.003 R4 8 19,191 34,039 -

64 0.003 R5 12 20,160 35,483 -

'84 0.003 R6 4 19,885 36,212 -

84 ,0.003 R7 4 18,431 35,663 -

64 0.003 S1 4 0 11,804 -

0 -

S2 4 0 17,117 -

28 0.003 S3 6 0 20,606 -

48 0.003 ,

S6 10 0 17,935 -

28 0.003 l l

S7 9 0 19,986 -

48 0.003 S8 8 0 21,027 -

64 0.003  !

Assumes EOC14=13,500 MWD/mU Assumes EOC15=14,200 MWD /WU Page 8 of 67  !

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

Hf Hf 3 4 5 6 7 P/ S1 R4 S6 R2 8 9 10 11 12 13 N/ R5 R2 S3 P S3 Hf 14 15 16 17 18 19 R1 R2 S6 P/ S7 R3 20 21 22 23 24 25 26 R2 S7 R5 R7 R5 88 P/ 27 28 29 30 31' 32 S2 P S8 R4 P/ R6 33 N

34 35 36 37 38 39 Q R2 S6 R3 S8 R6 S7 l

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Cycle 15 Core Loading Pattern Omaha Public Power District Figure 1

-Fort Calhoun Station Unit No.1 1'-

l. . Page 9 of 67 1

00000000000000 09000000000080 OO 000000 OO OO OOOOOO OO 00009000090000 00000000000000 000000 000000 000000 000000 .

00000000000000 00000000090000 '

OO 000000 OO OO 000000 OO 09000000000080 '

00000000000000 S - Shim (B C) 4 Rod (8)

O - Fuel Rod (168)

- Guide Tube Batch N/ Assembly- Omaha Public Power District Figure Fuel Rod and Poison Rod Locations Fort Calhoun Station Unit No.-1 3-2 Page 10 of 67

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)

- Guide Tube Batches N and P Fuel Rod Omaha Public Power District- Figure Locations Fort Calhoun Station Unit No.1 3-3 Page 11 of 67

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OOOOOOOOOOOOOO O9OOOOOOOOOOSO ,

OO OOOOOO OO OO OOOOOO OO 00000000090000 OOOOOOOOOOOOOO OOOOOO OOOOOO OOOOOO OOOOOO OOOOOOOOOOOOOO OOOO9OOOO9OOOO OO OOOOOO OO OO 000000 OO OOOOOOOOOOOO9O OOOOOOOOOOOOOO 9 - Shim (B 4C) Rod (8)

O - Low (3.25 W/o) Enrichment Fuel Rod (88)

O - High (3.95 W/a) 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 67.-

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k 00000000000000 ,

00000000000000 OO 000000 OO OO 000000 OO 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00000000000000 OO 000000 OO OO 000000 OO 00000000000000 OOOOOOOOOOOOOO O - Natural Uranium Fuel Rods (176)

- Guide Tube l

l Batch R1 Assembly Omaha Public Power District Figure Fuel Rod Locations Fort Calhoun Station Unit No.1 3-5 Page 13 of 67

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l 00000000000000 00000000000000 OO 000000 00 00 000000 00 00000000000000 00000000000000 000000 000000 000000 000000 00000000000000 00000000000000 OO 000000 00 OO 000000 OO 00000000000000 00000000000000 O - Low Enrichment Fuel Rod (52)

O - High Enrichment Fuel Rod (124)

- Guide Tube Batch Enrich ent (w/o) Enrichm!nt (w/o)

S1 3.50 4.00 Batch S1 Assembly Omaha Public Power District Figure Fuel Rod Locations Fort Calhoun Station Unit No.1 3-6 Page 14 of 67 '

00000000000000 ,

0000000000000O OO 000000 00 00 000000 00 00000000000000 00000000000000 000000 000000 000000 000000 0.0OOOOOOOOOOOO 00000000000000 OO 000000 00 00 000~O00 00 ,

00000000000000 00000000000000-O - Low Enrichment Fuel Rod (36)

O - Low Enrichment Fuel Rod with IFBA (16)

O - High Enrichment Fuel Rod (112)

O - High Enrichment Fuel Rod with IFBA (12) l

- Guide Tube Low High Batch Enrichment (w/o) Enrichment (w/o)

R2 3.50 4.00

. S2 - 3.50 4.00 S6 3.00 3.50 Batches R2, S2 and S6 Assembly Omaha Public Power District Figure -

Fuel Rod and 28 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-7 Page 15 of 67-

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00000000000000 OO OOOOOO OO OO 000000 OO '

00000000000000 00000000000000 000000 000000  !

000000 000000 00000000000000 00000000000000 ,

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OO-OO OOOOOO OO 000000000000L00 00000000000000 O - Low Enrichment Fuel Rod (20)

Q - Low Enrichment Fuel Rod with IFBA (32)-

O - High Enrichment Fuel Rod (10'8)

Q - High Enrichment Fuel Rod with IFBA (16)

- Guide Tube l

Low High Batch Enrichment (w/c) Enrichment (w/o)

.R3 3.50 4.00 S3 3.50 4.00 i S7 3.00 3.50 Batches R3, S3 and S7 Assembly Omaha Public Power District Figure Fuel Rod and 48 IFBA Rod Locations Fort Calhoun Station Unit No.1 3-8 i l

Page 16 of 67 j 1

00000000000000 00000000000000 OO 000000 00 00 000000 00 00000000000000 00000000000000 000.000 .000000 000000 000000 00000000000000 00000000000000 00 000000 00 -

00 000000. 0 0-00000000000000 00000000000000 ,

O - Low Enrichment Fuel Rod (12) -

O - Low Enrichment Fuel Rod with IFBA (40)

O - High Enrichment Fuel Rod (100) ,

O - High Enrichment Fuel Rod with IFBA (24)

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

R4 3.50 4.00 R7 3.25 3.75-

'S8 3.00 3.50 Batches R4, R7 and S8 Assembly- Omaha Public Power District Figure Fuel Rod and 64 IFBA Rod Locations Fort Calhoun Station Unit No.1 :3-9'  :

Page 17 of 67  :

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00000000000000 00000000000000 OO 00000O OO OO 000000 OO  ;

00000000000000 00000000000000 ,

O00000 0.00000 i O00000 000000 00000.000000000 1 0O000000000000 00 000000 00 OO 000000 OO 00000000000000-00000000000000 O - Low Enrichment Fuel Rod (12)

O - Low Enrichment Fuel Rod with IFBA (40)

O - High Enrichment Fuel Rod (80)

O - High Enrichment Fuel Rod with IFBA (44) .

- Guide Tube Low High Batch Enrichment (w/o) Enrichment (w/o)

!R5- 3.50 4.00 R6 3.25 3.75 Batches R5 and R6 Assembly Fuel Omaha Public Power District Figure Rod and 84 IFBA Rod Locations Fon Calhoun Station Unit No.1 3-10

' Page 18 of 67 - ,

2 AA - Assembly Location BB - Fuel Type C.CC - Initial Enrichment (w/o U-235)

DD,DDD - Assembly Average Exposure (MWD /MTU) 1 2 N N/.

3.70 3.70 27,612 33,689 3 4 5 6 7 P/ S1 R4 S6 R2 3.59 3.85 3.85 3.35 3.85 37,784 0 18,277 0 15,228 8 9 10 11 12 13 N/ R5 R2 S3 P S3 3.70 3.85 3.85 3.85 3.94 3.85 39,120 20,309 13,519 0 30,434 0 14 15 16 17 18 19 R1 R2 S6 P/ S7 R3 0.74 3.85 3.35 3.59 3.35 3.85 4,222 11,492 0 27,918 0 18,929 2 ", 21 22 23 24 25 R2 S7- R5 R7 RS- S8 26 3.85 3.35 3.85 3.60 3.85 3.35 p/ 13,482 0 20,040 18,431 20,130 0 3.59 27 28 29 30 31 32 37,875 S2 S8 R4 P P/ R6 33 3.85 3.94 3.35 3.85 3.59 3.60 N 0 29,083 0 20,105 27,351 19,833 3.70 34 35 36 37 38 39 29,490 Q1, R2 3.85 S6 3.35 R3 3.85 S8 3.35 R6 3.60 S7 3.35 14,630 0 19,592 0 19,938 0 Note: EOC 14 Core Average Burnup = 13,500 MWD /MTU l Cycle 15 BOC Initial Enrichment Omaha Public Power District Figure and Assembly Average Exposure Fort Calhoun Station Unit No.1 3 -11 Page 19 of 67

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DD,DDD l- Assembly Average Exposure (MWD /MTU)

N

• N/

3.70 3.70 30,333 36,763 P/ ^ S1 R4 S6 R2 3.59 3.85 3.85 3.35 3.85

• 41,823 11,804 30,040 13,985 28,277 8 9 10 11 12 13 N/ R5 R2 S3 P S3 3.70 3.85 3.85 3.85 3.94 3.85 41,308 30,180 30,054 20,416 46,047 20,985 14 15 16 17 18 - 19 R1 R2 S6 P/ S7 R3 0.74 3.85 3.35 3.59- 3.35 -3.85 8,751 27,129 20,321 44,194 21,155 38,432 I 20 21 22 23 24 25 R2 S7 .R5 R7 RS- S8 26 3.85 3.35 3.85 3.60 3.85 3.35' p/ 25,582 19,059- 38,329 35,663 37,941 20,742 3.59 27 28 29 30 31 32 42,069 S2 S8 R4 P P/ R6 33 3.85 3.94 3.35 S.85 3.59 3.60 N 17 117- 45,459 21,254 -38,039 42,405- 36,108 3.70 34 35 36 37- 38 39 35 471 -R2 S6 R3 S8 R6 S7 i

3.85 3.35 3.85 3,35- 3.60 3.35-39,295- 21,064 39,093 20,859 36,316 19,017 L

F Note: EOC 15 Core Average Burnup = 14,200 MWD /MTU Cycle 15 EOC Initial Enrichment - Omaha Public Power District Figure and Assembly Average Exposure ~

Fort Calhoun Station Unit No.-1 3-12.

w j-j Page 20 'of 67

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4.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch S fuelis the same as for the Batch R fuel supplied by Westinghouse (W) in Cycle 14.

The Batch S fuel is similar in design to the fuel supplied by ABB-Combustion Engineering (ABB-CE) and is mechanically, thermally, and hydraulically compatible with the ABB-CE fuel remaining in the Cycle 15 core. References 4 and 5 describe ABB-CE and W fuel characteristics and design, respectively. The Westinghouse fuel will not be resident in the reactor with any of the Exxon (i.e., Siemens) fuel previously used at Fort Calhoun Station.

Page 21 of 67

5.0 NUCLEAR DESIGN 5.1 PHYSICAL CHARACTERISTICS 5.1.1 Fuel Management The Cycle 15 fuel management uses an extreme low radial leakage design, with twice and thrice burned assemblies predominantly loaded on the periphery of the core with hafnium flux suppression rods inserted into the guide tubes of selected peripheral fuel assemblies adjacent to the reactor vessel limiting welds. Thic extreme low radial leakage fuel loading pattern is utilized to minimize 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 15 have been reduced by 'owering the enrichment of the fuel pins a'Jjacent to the fuel assembly water holes as described in Section 3.0.

Also described in Section 3.0 is the Cycle 15 loading pattern which is composed of 41 fresh Batch S assemblies, 37 of which contain the aforementioned IFBA pellet design. All of these 41 assemblies employ intra-assembly uranium enrichment splits. Batches S1 through S3 contain a high pin U-235 enrichment of 4.00 w/o and a low pin U-235 enrichment of 3.50 w/o, while batches S6 through S8 contain a high pin U-235 enrichment of 3.50 w/o and a low pin U-235 enrichment of 3.00 w/o. Sixteen thrice burned N assemblies are being returned to the core, along with 24 twice burned P assemblies and 52 once burned R assemblies. Four of the Batch R assemblies contain fuel rods that are loaded with naturally enriched uranium and also placed in locations near the lirriting reactor vessel welds. This assembly arrangement will produce a Cycle 15 loading pattern with a cycle energy of 13,500 MWD /MTU with an additional 700 MWD /MTU of energy in a coastdown mode if required. The Cycle 15 core characteristics have been examined for a Cycle 14 termination of 14,200 MWD /MTU and limiting values '

established for the safety analysis.

Physics characteristics including reactivity coefficients for Cycle 15 are listed in Table 5-1 along with the corresponding values from Cycle 14. 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 conditions for all events produce the minimum /most limiting scram worth and the smallest scram worth excess margin beyond the 4.0%Ap Technical Specification requirement. Fcr Cycle 15 this scram worth margin is 0.91%Ap.

Page 22 of 67

)

5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.1 Fuel Management (Continued)

L Table 5-2 presents a summary of the BOC, HZP and EOC, HZP CEA shutdown worths and reactivity allowances for Cycle 15. Because the  ;

Cycle 15 CEA worth values, used in the calculation of minimum scram worth, exceed the minimum value required by Technical Specifications, adequate shutdown margin is provided.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC15, MOC15, and EOC15, respectively, and are based upon the Cycle 14 late shutdown window burnup of 13,500  :

MWD /MTU. These relative power densities are. assembly averages-representative of the entire core length. The high burnup end of the Cvcle 14 shutdown window tends to increase the power peaking in the high power assemblies in the Cycle 15 fuel loading pattern. The radial power  :

distributions, with Bank 4 fully inserted at beginning and end of Cycle 15, 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 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 i peaking values actually used are higher than those expected to occur at any time during Cycle 15. 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.

i As previously indicated, Figures 3-5 and 3-6 show the integrated assembly burnup values at 0 and 14,200 MWD /MTU for Cycle 15. t

.{

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 shapes. The maximum three-dimensional or total peaking factor (Fq) anticipated in Cycle 15 during normal base load, all rods cut operation at full power is 2.0783, including uncertainty allowances.

t Page 23 of 67

[

5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.1 Fuel Management (Continued)

Table 5-2 presents a summary of the BOC, HZP and EOC, HZP CEA shutdown worths and reactivity allowances for Cycle 15. Because the Cycle 15 CEA worth values, used in the calculation of minimum scram worth, exceed the minimum value required by Technical Specifications, ,

adequate shutdown margin is provided.

5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial ,

power distributions at BOC15, MOC15, and EOC15, respectively, and are  !

based upon the Cycle 14 late shutdown window burnup of 13,500 MWD /MTU. These relative power densities are assembly averages representative of the entire ccre length. The high burnup end of the Cycle 14 shutdown window tends to increase the power peaking in the high power assemblies in the Cycle 15 fuel loading patterri. The radial power distributions, with Bank 4 fully inserted at beginning and end of Cycle 15, are shown in Figures 5-4 and 5-5, respectively.

I The radial power distributions described in this section are calcu!ated data without uncertainties o, 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 pecking values actually used are higher than those expected to occur at any time during Cycle 15. 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 burnup values at 0 and 14,200 MWD /MTU for Cycle 15.

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 shapes. The maximum three-dimensional or total peaking ,

factor (Fq) anticipated in Cycle 15 during normal base load, ARO operation at full power is 2.0783, including uncertainty allowances.

i l

i l

Page 23 of 67

-5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data Bounding reactiWy 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 15. These factors were based on the most limiting peaking factor for Cycle 14 and raised to a more bounding value to encompass future fuel cycles. As such the Table 5-3 values are projected to encompass the most limiting operating conditions anticipated during Cycles 14 through 18. The values shown were verified to bound actual Cycle 15 values which were calculated in accordance with Reference 3.

In addition, Table 5-3 also lists these values used from Cycle 14.  !

5.1.3.2 Dropped CEA Data The Cycle 15 safety related data for the dropped CEA analysis were calculated identically with the methods used in Cycle 14.

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 14.

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 15 are the same as those presented in Reference 2.

-)

Page 24 of 67 l H

j

TABLE 5-1 FORT CALHOUN UNIT NO.1, CYCLE 15 NOMINAL PHYSICS CHARACTERISTICS Units Cycle 14 Cycle 15 Critical Boron Concentration Hot Full Power, ARO, Equilibrium Xenon, BOC ppm 835 978 inverse Boron Worth Hot Full Power, BOC ppm /%Ap 113 117 Hot Full Power, EOC ppm /%Ap 90 91 Reactivity Coefficients with All CEAs Withdrawn Moderator Temperature Coefficient (MTC)

Beginning of Cycle, HZP 10-'Ap/ F + 0.09 + 0.34 End of Cycle, HFP 10~'Ap/ F - 2.80 -2.84 Doppler Coefficient (FTC)

. Hot Full Power, BOC 10"Ap/ F - 1.51 -1.60 Hot Full Power, EOC 10"Ap/ F -1.69 - 1.77 -

Total Delayed Neutron Fraction, (L Hot Full Power, BOC 0.00625 .0.00620 Hot Full Power, EOC 0.00518 0.00512 Neutron Generation Time, l*

Hot Full Power, BOC 10"sec 21.6 20.8 Hot Full Power, EOC 10"sec 27.2 26.8 4

Page 25 of 67

TABLE 5-2 FORT CALHOUN UNIT NO.1, CYCLE 15 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT ZERO POWER BOC, HZP EOC, HZP ,

(%Ap) (%Ap)

1. Worth of all CEAs Inserted 7.45 9.01. l

2. Stuck CEA Allowance 1.34 1.97- ,

3. Worth of all CEAs Less Worth of Most Reactive CEA Stuck Out 6.11 7.04

4. Power Dependent Ir'sertion Limit CEA Worth 1.10 1.14

5. Calculated Scram Worth 5.01 5.90

6. Physics Uncertainty plus Bios 0.10 0.12

-7. Net Available Scram Worth 4.91 5.78 *

8. Technical Specification -

Shutdown Margin 4.00 4.00

9. Margin in Excess of Technical Specification Shutdown Margin 0.91 1.78 a

t Page 26 of 67 -

]

1 TABLE 5-3 FORT CALHOUN UNIT NO,1, CYCLE 15 i BOUN< ) LNG CEA EJECTION DATA- j l

Maximum Radial . .

Power Peaking Factor BOC 14 Value EOC 14 Value BOC 15 Value EOC 15 Vah>e Full Power with Bank 4 inserted; worst CEA ,

ejected 3.73 3.73 3.73 3.73 Zero Power with Banks 4+3 inserted:

worst CEA ejected ' 5.74 5.74 5.74 5.74 I Maximum Ejected CEA Worth (%Ap)

Full Power with Bank 4 inserted worst CEA ejected 0.36 0.36 0.36 0.36 Zero Power with Banks 4+3 inserted.

worst CEA ejected 0.69 0.69 0.69 0.69 i

NOTE: The most limiting radial peaking factor was calculated (including uncertainties and biases) for the BOC and EOC conditions. The peaking factors were then raised to a ,

more limiting value, and the largest value was transmitted to Westinghouse. This value was then applied in a conservative manner to operating conditions during a cycle to ensure the existing and future operating cycles would be bounded by the Westinghouse CEA ejection analysis.

4

' Page 27 of 67 l o

. . . = . .

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

1 2 0.1652 0.1834 3 4 5 6 7 0.2765 0.8793 0.8301 0.9219 0.8757 8 9 10 11 12 13 0.1385 0.7184 1.2595 1.4287 1.0739 1.4056 14 15 16 17 18 19 0.2742 1.1497 1.5210 1.1914 1.4491 1.3998 1.7674 20 21 22 23 24 25 0.8287 1.2826 1.3623 1.3173 1.3217 1.3749 26 0.2647 27 28 29 30 31 32 1.1996 1.1437 1.3'96 1.3166 1.0852 1.1876 33 0.3879 34 35 36 37 38 39 C _

,.,4,6 ,.5,08 1.3897 1.367, ,.,940 ,3075 Maximum 1-Pin Peak at 23% Core Height Cycle.15 Assembly Power Distribution Omaha Public Power District . Figure 0 MWD /MTU, HFP, Equilibrium Xenon Fort Calhoun Station Unit No.1 5-1 Page 28 of 67 ,

~

AA - Assembly Location B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.1855 0.2106 e

3 4 5 6 7 0.2710 0.8212 0.8271 0.9942 0.9280 8 9 10 11 12 13 0.1465 0.6819 1.1530 1.4531 1.0883 1.5121 1.7405 14 15 16 17 18 19 0.3158 1.0944 1.4281 1.1242 1.5219 1.3961 20 21 22 23 24 25 0.8520 1.3637 1.2923 1.2120 1 2607 1.5026 26

27. 28 29 30 31 32 1.2075 1.1392 1.5432 1.2750 1.0409 1.1530 33 34 35 36 37 38 39 1.1004 1.4992 1.3986 1.5155 1.1612 1.3581-Maximum 1-Pin Peak at 23% Core Height Cycle 15 Assembly Power Distribution Omaha Public Power District Figure l 7,000 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-2 i

Page 29 of 67 -]

AA - Assembly Location B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.2264 0.2575 1

3 4 5 6 7 0.3317 0.9019 0.8919 1.0730 0.9906 8 9 10 11 12 13 0.1870 0.7622 1.1773 1.4474 1.0856 1.4737 1.6467 14 15 16 17 18 19 0.4003 1.1311 1.3864 1.0951 1.4286 1.2972 20 21 22 23 24 25 0.9237 1.3586 1.2319 1.1531 1.1866 1.4042 26 27 28 29 30 31 32 1.2569 1.1148 1.4391 1.1939 1.0089 1.1091 33 0.4827 34 35 36 37 38 39 1.1303 1.4178 1.2866 1.4071 1.1131 1.3100 Maximum 1 -Pin Peak at 23% Core Height  !

l l

l Cycle 15 Assembly Power Distribution Omaha Public Power District Figure ,

14,200 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-3 i

Page 30 of 67  ;

AA - Assembly Location i B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.1735 0.1959 3 4 5 6 7 0.2146 0.8199 0.8411 0.9723 0.9340 8 9j 10 11 12 13 0.1003 j0M072; 1.1625 1.4484 1.1278 1.4899 14 15 16 17 18 19 0.2589 1.0668 1.4933 1.2206 1.5114 1.4655 20 21 22 23 24 25 0.8672 1.3227 1.4034 1.3577 1.3469 1.3878 26 0.2869 27 28 29 30 31 32 1.2928 1.2203 1.4704 1.3451 1.0336 1.0468 33 0.4243 34 35 36 37 38 391 1.2393 1.6235 1.4677 1.3844 1.0541 ;i0.7344?

1.8823 Maximum 1-Pin Peak at 17% Core Height

- Bank 4 Locations Cycle 15 Assembly RPD Bank 4 in Omaha Public Power District Figure 0 MWD /MTU, HFP, Equilibrium Xenon Fort Calhoun Station Unit No.1 5-4 Page 31 of 67

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

0.2410 0.2785  ;

3 4 5 6 7 0.2491 0.8374 0.9128 1.1437 1.0673 8 94 10 11 12 13 0.1330 (0A058) 1.0758 1.4773 1.1501 1.5750 1.7501 14 15 16 17 18 19 0.3772 1.0395 1.3583 1.1277 1.4992 1.3663 20 21 22 23 24 25 .

0.9685 1.4047 1.2741 1.1929 1.2106 1.4172 26 27 28 29 30 31 32 1.3598 1.1937 1.5182 1.2208 0.9485 0.9524 33 0.5313 34 35 36 37 38 396 .

1.2332 1.5297 1.3626 1.4238 0.9574 EU[G$d23 Maximum 1-Pin Peak at 17% Core Height

- Bank 4 Locations l l

1 Cycle 15 Assembly RPD Bank 4 In Omaha Public Power District Figure 'l 14,200 MWD /MTU, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-5 i l

Page 32 of 67

6.0 THERMAL-HYDRAULIC DESIGN i

6.1 DNBR ANALYSIS i Steady state DNBR analyses of Cycle 15 at the rated power of 1500 MWt have  ;

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, and was combined with the use of the TORC code for transient DNBR analyses. The DNBR analysis applications and methods did not change from the previous cycle. In these analyses the TORC computer code was used to calculate the minimum DNBR. Both codes are approved for use with the OPPD methods. The reload methodology for Cycle 15 can be found in Reference 5.

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 factor, engineering iactor on hot channel heat input, rod itch and clad diameter factor) listed in Table l 6-1 have been combined statistically with other uncertainty factors at the 95/95  !

confidence / probability level (Reference 6) 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 for A38-CE fuel at 45,000 MWD /MTU burnup is 0.5% in MDNBR. This penalty was applied in the derivation of the statistical combination of uncertainties (SCU) (Reference 6) MDNBR design limit of 1.18 (References 4 and 8). Westinghouse has identified in the fuel mechanical design report that the amount of deflection does not require a DNB penalty to be applied under Westinghouse analysis requirements. To simplify the assumption for the amount of fuel rod bow allowed in the Westinghouse fuel and for the ABB-CE fuel the design basis is the same. Thus, the ABB-CE DNB penalty was applied to the Westinghouse fuel to ensure that the OPPD statistical combination of uncertainties were still valid and that conservative input assumptions were used in the analysis.

Page 33 of 67

- 6.0 THERMAL-HYDRAULIC DESIGN ,

6.1 DNBR ANALYSIS Steady state DNBR analyses of Cycle 15 at the rated power of 1500 MWt have 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, and l was combined with the use of the TORC code for transient DNBR analyses. The DNBR analysis applications and methods did not change from the previous cycle in these analyses the TORC computer code was used to calculate the minimum DNBR. Both codes are approved for use with the OPPD methods. The reload methodology for Cycle 15 can be found in Reference 5.

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 factor, engineering iactor 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 / probability level (Reference 6) 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 for ABB'-CE fuel at 45,000  :

MWD /MTU burnup is 0.5% in MDNBR. This penalty was applied in the derivation of the statistical combination of uncertainties (SCU) (Reference 6) MDNBR design limit of 1.18 (References 4 and 8). Westinghouse has identified in the fuel mechanical design report that the amount of deflection does not require a DNB penalty to be applied under Westinghouse analysis requirements. To simplify the assumption for the amount of fuel rod bow allowed in the Westinghouse fuel and for the ABB-CE fuel the design basis is the same. Thus, the ABB-CE DNB penalty was applied to the Westinghouse fuel to ensure that the OPPD statistical combination of uncertainties was still valid and that conservative input '

assumptions were used in the analysis.

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Page 33 of 67-i

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TABLE 6-1 FORT CALHOUN UNIT NO.1, CYCLE 15 i THERMAL HYDRAULIC PARAMETERS AT FULL POWER  ;

Unit Cycle 15*

Total Heat Output (Core Only) MWt 1500 l 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 Inlet Temperature (Maximum) F 543' >

Total Reactor Coolant Flow gpm 202,500 (Steady State) 10' Ibm /hr 76.52-(Through the Core) 10' Ibm /hr 73.25 Hydraulic Diameter (Nominal Channel) ft .044 q Average Mass Velocity 10' Ibm /hr-ft 2.2314

{

Core Average Heat Flux (Accounts for Heat Generated BTU /hr-ft' 179469 in Fuel Rod)

E Total Heat Transfer Surface Area ft' 28,526"*

Average Core Enthalpy Rise BTU /lbm 72.6 Average Unear Heat Rate kW/ft 5.95*

• 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 intet temperature and nominal primary system pressure were used to -

calculate these parameters.  :

• • Based on Cycle 15 specific value of 192 fuel displacing shims.

• • • These factors were combined statisticsily (Reference 7) with other uncertainty factors at 95/95 confidence /probabilit'y level to define a design limit on CE minimum DNBR.

Page 34 of 67  !

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7.0 TRANSIENT ANALYSIS l

This section presents the results of the Omaha Public Power District Fort Calhoun  !

Station Unit 1, Cycle 15 Non-LOCA safety analyses at 1500 MW1.

l The Design Bases Events (DBEs) considered in the safety analysis are listed in Table j 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 limits. -

2. AOOs for which the initial steady state thermal margin, maintained 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 of (i.e., conservative with respect to) the reference cycle values (Fort Calhoun Updated Safety Analysis Report including Cycle 14 analyses, Reference 1). For these DBEs, the results and conclusions quoted in the reference cycle analysis remain valid for Cycle 15.

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 15 results would be bounded by Cycle 14 or the USAR reference cycle.

Events have been evaluated for up to a total of 6% steam generator tube plugging, where conservative, since Cycle 11. Fort Calhoun Station currently has 1.08% steam generator tubes plugged; thus, no additional analysis is required.

For each of the events reanalyzed, Table 7-3 shows the reason for the reanalysis, the acceptance criteria 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 35 of 67

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l 7.0 TRANSIENT ANALYSIS l i

This section presents the results of the Omaha Public Power District Fort Calhoun 3 Station Unit 1, Cycle 15 Non-LOCA safety analyses at 1500 MWt.

The Design Bases Events (DBEs) considered in the safety analysis are listed in Table i 7-1. These events were categorized in the following groups: l

1. Anticipated Operational Occurrences (AOOs) for which the intervention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable -

limits.

2. AOOs for which the initial steady state thermal margin, maintained by Limiting Conditions for Operation (LCO), is 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 iruicated in Table 7-1, no reanalysis was performed for the DBEs for which key transient input parameters are within the bounds of (i.e., conservative with respect to) the reference cycle values (Fort Calhoun Updated Safety Analysis Report including  :

Cycle 14 analyses [ Reference 1]). For these DBEs, the results and conclusions quoted in the reference cycle analysis remain valid for Cycle 15.

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 15 results  ;

would be bounded by Cycle 14 or the USAR reference cycle. j Events have been evaluated for up to a total of 6% steam generator tube plugging, >

where conservative, since Cycle 11. Fort Calhoun Station currently has 1.08% steam ,

generator tubes plugged; thus, no additional analysis is required.

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

Page 35 of 67

r TABLE 7-1 ,

il FORT CALHOUN UNIT NO.1, CYCLE 15 .

DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS 7.1 Anticipated Operational Occurrences for which intervention of the RPSis necessary to prevent exceeding acceptable limits:

7.1.1 Reactor Coolant System Depressurization Reanalyzed 7.1.2 Loss of Load Not Reanalyzeds 7.1.3 Loss of Feedwater Flow Not Reanalyzed 5 7.1.4 Excess Heat Removal due to Feedwater Malfunction Not Reanalyzeds-7.1.5 Startup of an inactive Reactor Coolant Pump Not Reanalyzed 1 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 Reviewed3,5 7.2.4 - CEA Drop Reanalyzed 7.2.5 Boron Dilution Reanalyzed 7.2.6 Transients Resulting from the Malfunction ,

of One Steam Generator Not Reanalyzed 4 )

7.3 Postulated Accidents .)

i 7.3.1 CEA Ejection Not Reanalyzeds -

7.3.2 Steam Line Break Reviewed 5 7.3.3 Seized Rotor Reviewed 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. l 3 Requires Low Flow Trip.

4 Requires trip on high differential steam generator pressure. i 5 Event bounded by reference cycle analysis. A negative determination utilizing the_10 CFR 50.59 criteria was made for this event.

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Page 36 of 67

.. J TABLE 7-2 l FORT CALHOUN UNIT NO.1, CYCLE 15  ;

. CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Physics Parameters Units Cycle 14 Values Cycle 15 Values -

Radial Peaking Factors For DNB Margin Analyses .

(F;)

Unrodded Region 1.79* 1.77*

Bank 4 Inserted 1.92* 1.89*

l For Planar Radial Com.conent (FL ) of 3-D Peak (CTM Limit Analyses)

Unrodded Region 1.85* 1.86*

Bank 4 Inserted 2.0* 1.98*

Maximum Augmentation Factor 1.000 1.000 ,

ModeratorTemperature Coefficient 10-4 Ap/ F -3.0 to +0.5 -3.C 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.

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1 Page 37 of 67 3

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r TABLE 7-2  !

(Continued)

I Safety Parameters Units Cycle 14 Values Cycle 15 Values 1

Power Level MWt 1500*- 1500*

Maximum Steady State Temperature F 545* 543* i Minimum Steady State  :

Pressurizer Pressure psia 2075* 2075* .

l Reactor Coolant Flow gpm 202,500* 202,500* i 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 Bank 4 25 25 .:

a

' Maximum Initial Linear i Heat Rate for Transient  ;

Other than LOCA. kW/ft 13.8. 15.5

!~

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 P

Minimum DNBR (CE-1) 1.18* 1.18* l The effects of uncertainties on these parameters were accounted for statistically J in the DNBR 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 38 of 67 l

1

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TABLE 7-3 +

FORT CALHOUN UNIT NO.1 DESIGN BASIS EVENTS REANALYZED FOR CYCLE 15 t Reason for Acceptance Summary Event Reanalysis Criteria of Results Sequential CEA Calculate cycle specific Minimum DNBR 2 MDNBR =1.484 -l Group Withdrawal ROPM values 1.18 using the CE-1 PLHGR< 22 kW/ft correlation. Transient PLHGR g 22 kW/ft.

CEA Drop Calculate cycle specific Minimum DNBR 2 MDNBR = 1.467 - -

ROPM values 1.18 using CE-1 PLHGR < 22 kW/ft  ;

correlation. Transient. l PLHGR S 22 kW/ft i

Excess Load Calculate cycle specific Minimum DNBR 2 MDNBR = 1.436 i ROPM values 1.18 using CE-1 PLHGR < 22 kW/ft correlation. Transient -l PLHGR g 22 kW/ft j

RCS Depressurization To provide a conservative Pb:asvalue g the Poias = 30 psia  :

Pbias input for the TM/LP due previous cycle's limiting to the Excess Load value methodology change Boron Dilution Evaluate the effects of Operator action times are All available times for >

changes in the inverse Boron large enough to terminate operator action are j Worth and Critical Boron the event prior to violating greater than the  !

Concentrations SAFDL limits . requirements. .

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Page 39 of 67

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I 7.0 TRANSIENT ANALYSIS (Continued) 7.1 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 1) 7.1.1 RCS Depressurization Event i The RCS Depressurization event was reanalyzed for Cycle 15 to l determine the pressure bias term for the TM/LP trip setpoint.

The RCS Depressurization event is the Design Basis Event analyzed to determine the maximum pressure bias term input to the TM/LP trip. The methodology used for Cycle 15 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 transien' are moni'rred by the TM/LP trip directly. Consequently, with TM/LP trip ,

setpu.nts 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 listing of the key input parameters while Table 7.1.1-2 ,

summarizes the 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 irip 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 determines the pressure bias term.

Page 40 of 67

TABLE 7.1.1 -1  ;

FORT CALHOUN UNIT NO.1, CYCLE 15 .

l

' KEY PARAMETERS ASSUMED IN THE RCS DEPRESSURIZATION ANALYSIS .

Parameter Units Cycle 15 i Initial Core Power Level MWt 1530 Core inlet Coolant Temperature F 545  ;

Pressurizer Pressure psia 2172 ]

Moderator Temperature Coefficient x 10-4 Ap/ F -3.0 Fuel Temperature Coefficient x 10-4 Ap/ F Most negative i predicted during -

core life t 2

Core Average Hgap BTU /hr-ft -F 500 s Total Trip Delay Time sec 1.4 TABLE 7.1.1 -2 FORT CALHOUN UNIT NO.1,' CYCLE 15 SEQUENCE OF EVENTS FOR RCS DEPRESSURIZATION ,

Time (sec) Event Setpoint'or Value 0.00 Inadvertent Opening of Both ------

Pressurizer Power Operated .,

Relief Valves  !

6.92 Reactor Trip 2078.89 psia 1

8.62 Time of Minimum DNBR 2052.16 psia i

')

H Page 41 of 67

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) 7.2.1 Excess Load Event The Excess Load event is protected by an RPS trip and maintenance of sufficient initial thermal margin as required by the LCOs.

The Excess Load event was reanalyzed for Cycle 15 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 analysis is described in Reference 6. The key input parameters used in the Cycle 15 Excess Load analysis are prt,sented 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 decalibration of the excore detectors due to RCS cooldown. Also, the time -

constants for the hot 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) RO PMs to input to the calculation 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.

Table 7.2.1-2 summarizes the sequence of events for the Excess Load analysis. It is concluded that the Excess Load event is the most limiting of the AOOs dependent upon initial overpower margin. When initiated from the Technical Specification LOOS, the event will not result in exceeding the DNBR or CTM design limits.

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Page.42 of 67

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TABLE 7-.2.1 -1 FORT CALHOUN UNIT NO.1, CYCLE 15 KEY PARAMETERS ASSUMED IN THE EXCESS LOAD ANALYSIS Parameter Units Cycle 15 Iritial Core Power Level MWt 1500* ,

Core inlet Coolant Temperature F 543*

Pressurizer Pressure psia 2075*

Moderator Temperature l Coefficient x 10-4 Ap/'F -0.846 Doppler Coefficient Multiplier 0.85 CEA Worth at Trip %Ap 5.7922 Cold Leg RTD Time Constant sec 12.0 .

Hot Leg RTD Time Constant sec 3.0

• The DNBR calculations used the methods discussed in Section 6.1 of this document and l detailed in References 2 through 5. The effects of uncertainties on these parameters were 1 accounted for statistically in the DNBR and CTM calculations.

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Page 43 of 67

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> TABLE 7.2.1 -2 1 FORT CALHOUN UNIT NO.1, CYCLE 15 l SEQUENCE OF EVENTS FOR THE EXCESS LOAD ANALYSIS j l

p Time (sec) Event Setpoint or Value 0.00 Inadvertent Opening of Steam ------ 1 Dump and Bypass Valves 72.10 High Power Trip Conditions 112% of Rated Power Reached 72.60 High Power Trip Signal 112% of Rated Power l

Generated 72.80 Tim)of Minimum DNBR 1.436 73.10 CEAs Begin to Drop into Core ------

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.i Page 44 of 67 L -

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

7.2.2 CEA Withdrawal Event The CEA Withdrawal (CEAW) event was reanalyzed for Cycle.15 to j determine the initial margins that must be maintained by the Limiting  ;

Condtions for Operationt. (LCOs) such that the DNBR and fuel centerline l 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- l sufficient initial steady state thermal margin provided by the DNBR and Linear Heat Rate (LHR) related LCOs. l The key input parameters used for the hot full power CEAW case are presented in Table 71' 2-1. For the HFP CEAW DNBR analysis, a MTC - l value approximately equalto 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.

Table 7.2.2-3 summarizes the sequence of events for the HFP.CEAW case. The HFP case for Cycle 15 requires less overpower margin (ROPM) than the ROPM for the CEA Drop analysis. The CEA Withdrawal ROPM '

requirements at lower powers are limiting for the transient events and are  ;

used as input inta t% setpoint analysis for establishment of the DNB LCO. ,

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 L,p, initiated by the Variable High Power Trip at 30% ( 20% plu's - -l 10% uncertainty of rated thermal power) was assumed in the analysis.

The key input parameters used for the zero power case are presented in ,

Table 7.2.2-2.  !

The 10 CFR 50.59 criteria are satisfied for the HZP event if the ' minimum - 3 DNBR is greater than that reported in the reference cycle. i Table 7.2.2-4 summarizes the sequence of events for the HZP'CEAW .

case. The zero power case initiated at the limiting conditions of operation results in a a minimum CE-1 DNBR of 5.44 which is less than the -

~

t reference cycle (Cycle 12) value of 6.99, but still far in excess of theu  !

minimum 1.18 DNBR limit. The analysis shows that the fuel to centerline j melt temperatures are well below those' corresponding to the acceptable j fuel to centerline melt limit.

Page 45.of 67 -i l

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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) event was reanalyzed for Cycle 15 to determine the initial margins that must be maintained by the Limiting Conditions for Operation (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 analyzino the CEAW event. This event is classified as one for which the accept.able DNBR and CTM limits are not violated by virtue o' maintenance of sufficient initial steady state thermal margin provided by the DNBR and Linear Heat Rate (LHR) related LCOs.

The key input parameters used for the hot full power CEAW case are presented in Table 7.2.2-1. For the HFP CEAW DNBR analysis, an MTC value approximately equal 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. ,

Table 7.2.2-3 summarizes the sequence of events for the HFP CEAW case. The HFP case for Cycle 15 requires less overpower margin (ROPM) -

than the ROPM for the CEA Drop analysis. The CEA Withdrawal ROPM requirements at lower powers are limiting for the transient events and are used as inputinto the setpoint analysis for emabllshment of the DNB LCO.

The zero power case was analyzed to der .att . thatacceptable DNBR and centerline melt limits are not exceeoc ! For i a zero power case, a reactor trip, initiated by the Variable High Funa i + 9130% ( 20% plus 10% uncertainty of rated thermal power) was assumed in the analysis.

The key input parameters used for the zero power case are presented in Table 7.2.2-2.

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.

Table 7.2.2-4 summarizes tr.e sequence of events for the HZP CEAW case. The zero power case initiated at the limiting conditions of operation results in a a minimum CE-1 DNBR of 5.44 which is less than the reference cycle (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.

Page 45 of 67

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

It is 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 is further concluded that the HFP initial available overpower margin requirements for this event were bounded by those of the Excess Load event.

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Page 46 of 67 I

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TABLE 7.2.2-1 FORT CALHOUN UNIT NO.1, CYCLE 15 1 KEY PARAMETERS ASSUMED IN THE HFP CEA WITHDRAWAL ANALYSIS Parameter Units Cycle 15 i

Initial Core Power Level MWt 1500*  ;

- Core Inlet Coolant Temperature F 513*

i Pressurizer Pressure psia 2075*

Moderator Temperature Coefficient x 10-4 Ap/*F 4 0.5 Doppler Coefficient Multiplier 0.85 Reactivity Inse tion Rate Range x 10-4 Ap/sec 0 to 3.0 '

CEA Group Withdrawal Rate in/ min 46 Holding Coil Delay Time sec 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 47 of 67

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.. q TABLE 7.2.2-2 FORT CALHOUN UNIT NO.1, CYCLE 15 KEY PARAMETERS ASSUMED IN THE HZP CEA WITHDRAWAL ANALYSIS i l

Parameter Units Cycle 12 Cycle 15

. initial Core Power Level MWt 1 1*

Core inlet Coolant Temperature F 532 532*

Pressurizer Pressure psia 2053 2075' r

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- 4.91 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.

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TABLE 7.2.2-3 FORT CALHOUN UNIT NO.1, CYCLE 15-  :

SEQUENCE OF EVENTS FOR THE HFP CEA WITHDRAWAL ANALYSIS Time (sec) Event Setpoint or Value ,

0.00 Inadvertent Withdrawal of ------

CEAs ,

626.67 Time of Minimum DNBR 1.484 788.94 Excore Power Approaches Trip Limit 112 %

TABLE 7.2.2-4 FORT CALHOUN UNIT NO.1, CYCLE 15 SEQUENCE OF EVENTS FOR THE HZP CEA WITHDRAWAL ANALYSIS Time (sec) Event Setpoint or Value 0.00 Inadvertent Withdrawal of ------

CEAs 16.57 High Power Trip 30 %

17.78 Peak Power Reached 71.09 % .,

18.56 Time of Minimum DNBR 5.857 l

Page 49 of 67

TABLE 7.2.3- 1 FORT CALHOUN UNIT NO.1, CYCLE 15 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Cycle 12 Cycle 15 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 10-4 Ap/'F + 0. 5 + 0.5 Doppler Temperature Multiplier 0.85 0.85 CEA Worth at Trip (ARO) %Ap - 6.50 - 6.54 LFT Analysis Setpoint  % of initial flow 93 93 LFT Response Time sec 0.65 0.65 CEA Holding Coil Delay 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 1.77 The uncertainties on these parameters were combir.ed statistically rather than deterministically. The values listed represent the bounds included in the statistical combination.

Page 51 of 67 1; r ly

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.4 Full Length CEA Drop Event The Full Length CEA Drop event was reanalyzed for Cycle 15 to determine the initial margins that must be maintained by the Limiting Conditions for Operation (LCOs) such that the DNBR and fuel CTM design limits will not be exceeded.

This event was analyzed parametrically in initial axial shape and rod configuration using the methods described in Reference 6. Table 7.2.4-1 lists the key input parameters used for Cycle 15 and cornpares them to the reference cycle (Cycle 11) values, while Table 7.2.4-2 contains a sequence of events for the CEA Drop analysis.

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 1.434. A maximum allowable initiallinear heat-generation rate of 18.4 kW/ft could exist as an initial condition without exceeding the acceptable fuel CTM limit of 22 kW/ft during this transient.

This amount oi margin is assured by setting the LHR related LCOs based on the more limiting allowable LOCA linear heat rate.

It is concluded that the CEA Drop event, when initiated from the Technical Specification LCOs, will not lead to a DNBR or fuel temperature which violates the DNBR and CTM design lirnits. It is further concluded that the HFP initial available overpower margin requirements for this event are bounded by that of the Excess Load event.

Page 52 of 67

TABLE 7.2.4-1 FORT CALHOUN UNIT NO.1, CYCLE 15 KEY PARAMETERS ASSUMED IN THE HFP CEA DROP ANALYSIS Parameter Units Cycle 11 Cycle 15 initial Core Power Level MWt 1500* 1500*

Core inlet Coolant Temperatore F 543* 543*

Pressurizer Pressure psia 2075* 2075*

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

Moderator Temperature Coefficient x 10-4 Ap/ F - 2.7 - 3.0 Doppler Coefficient Multiplier 1.15 1.45 CEA insertion at Maximum Allowed Power  % Insertion of Bank 4 25 25 Dropped CEA Worth Unrodded, %Ap - 0.2337 - 0.2892 PDIL, %Ap - 0.2295 - 0.2891-Maximum Allowed Power Shape Index at Negative Extreme of LCO Band - 0.18 -0.18 Radial Peaking Distortion .

Factor Unrodded Region 1.1566 1.1946 Bank 4 Inserted 1.1598 1.1945 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 53 of 67

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TABLE 7.2.4-2 FORT CALHOUN UNIT NO.1, CYCLE 15 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 i

1.1 Core Power Level Reaches a Minimum 65% of 1500 MWt and Begins to Return to Power Due to i Reactivity Feedbacks 88.7 Core inlet Temperature Reaches a 536.8 F ,

Minimum Value 200.0 RCS Pressure Reaches a Minimum 1999.2 psia Value 200.0 Core Power Returns to its Maximum 95% of 1500 MWt Value 200.0 Minimum DNBR is Reached 1.467 (CE-1 Correlation) j i

Page 54 of 67  :

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.5 Boron Dilution Event The Boron Dilution event was reanalyzed for Cycle 15 to verify that sufficient time is available for an operator to identify the cause and to terminate a boron dilution event for any mode of operation befors SAFDL limits are violated.

Table 7.2.5-1 compares the values of the key transient parameters-assumed in each mode of operation for Cycle 15 and the reference Cycle

14. Table 7.2.5-2 provides the results of the time to lose shutdown margin calculations for Cycle 14 and Cycle 15. The Cycle 15 analysis utilized a mass basis in the calculations, as was used in Cycle 14, rather than a volumetric basis to ensure that all operating temperature ranges for -

all modes of operation were bounded. The IBW value for hot standby was conservatively reduced this cycle to -55 from the previous cycle value of

-90.

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I 7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.5 Boron Dilution Event The Baron Dilution event was reanalyzed for Cycle 15 to verify that sufficient time is available for an operator to identify the cause and to terminate a boron dilution event for any mode of operation before SAFDL limits are violated.

Table 7.2.5-1 compares the values of the key transient parameters .

assumed in each mode of operation for Cycle 15 and the reference cycle, Cycle 14. Table 7.2.5-2 provides the results of the time to lose shutdown margin calculations for Cycle 14 and Cycle 15. The Cycle 15 analysis utilized a mass basis in the calculations, as was used in Cycle 14, rather than a volumetric basis to ensure that all operating temperature ranges for all modes of operation were bounded. The IBW value for hot standby was conservatively reduced this cycle to -55 from the previous cycle value of

-90.

.Page 55 of 67

TABLE 7.2.5-1 ,

FORT CALHOUN UNIT NO.1. CYCLE 15 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Parameters Cycle 14 Values Cycle 15 Values Critical Boron Concentration, ppm (ARO, No Xenon)

Mode Hot Standby 1292 1541 Hot Shutdown 1292 1541  !

Cold Shutdown - Normal RCS Volume 1204 1393 Cold Shutdown - Minimum RCS Volume

• 1204 1213 Refueling 1180 1358 Inverse Boron Worth, ppm /%Ap Mode Hot Standby -90 -55 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 Hot Standby - 4.0 -4.0 j 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)** - 1700 1900***

Shutdown Groups A and B out, all Regulating Groups inserted except most reactive rod =

stuck out.

Includes a 5.0%Ap shutdown margin.

• • • Proposed Cycle 15 COLR value.

Page 56 of 67

9 TABLE 7.2.5-2 FORT CALHOUN UNIT NO 1, CYCLE 15 RESULTS FROM THE BORON DILUTION ANALYSIS Time to Lose Prescribed Criterion for Minimum Time Mode Shutdown Margin (Min) to Lose Prescribed Shutdown Margin (Min)

Cycle 14 Cycle 15 Hot Standby 66.61 36.90 15 Hot Shutdown 42.61 36.79 15 Cold Shutdown 44.96 39.22 15 Normal RCS Volume Cold Shutdown 16.03 15.91 15 Minimum RCS Volume Refueling 59.44 41.89 30 t

Page 57 of 67

9 7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS

.3.1 CEA Ejection The CEA Ejection event was reviewed for Cycle 15 against the Westinghouse analysis supplied for Cycle 14 (Reference 14) with the new fuel design and operating parameters.

7.3.2 Steam Line Break Accident This accident was reviewed for Cycle 15 using the methodology discussed in References 6 and 12. The Steam Line Break (SLB) accident was previously analyzed in the Fort Calhoun FSAR and satisfactory results were reported therein. The SLB accidents at both HZP and HFP were examined in the reference cycle (Cycle 8) safety evaluation with acceptable results obtained. Both the FSAR and reference cycle evaluations are reported in the 1992 update of the Fort Calhoun Station Unit No.1 USAR.

The Full Power Steam Line Break accident was reviewed for Cycle 15 for a more negative MTC of -3.0 x 10-4 Ap/ F than the -2.5 x 10-4 Ap/ F -

value that was used in the Cycle 8 analysis. However, the cooldown curve for Cycle 15 is bounded by Cycle 8 (as shown in Figure 7.3.2-1). This _

figure shows that the reactivity insertion for the Cycle 15 core with an MTC of -3.0 x 10-4 Ap/ F due to a SLB accident at full pogver is substantially less than the value used in the Cycle 8 analysis. (This smaller reactivity insertion is due to the use of the DIT cross-sections which are valid for a range of moderator temperatures from room temperature to 600 K while the analyses prior to Cycle 9 were performed with cooldown curves derived by conservatively extrapolating CEPAK cross-section values to low temperatures.) The Cycle 15 minimum available shutdown worth at HFP is 6.2885 %Ap cor spared to a Cycle 8 value of 6.68%Ap. This implies a margin decrease u 0.395%Ap. The Cycle 15 moderator cooldown reactivity between 574 F and 350 F at HFP is 4.7%Ap compared to 5.37

%Ap in Cycle 8. This implies a margin increase of 0.67 %Ap. The Cycle 15 doppler coefficient is more negative than the Cycle 8 doppler coefficient including uncertainties. However, this loss in margin is offset by the gain in margin from the moderator cooldown reactivity. The net gain ensures that the overall reactivity insertion for a Cycle 15 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.

l Page 58 of 67

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- 7.0 TRANSIENT ANALYSIS (Continued) l 7.3 POSTULATED ACCIDENTS (Continued).

7.3.2 Steam Line Break Accident (Continued) i A similar evaluation was performed for the Hot Zero Power SLB accident. I

' Again, the Cycle 15 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). Since the. minimum available shutdown margin for Cycle 15 remains unchanged from the reference cycle value (4.0%Ap), the overall reactivity insertion for the Cycle 15 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 consequences 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 15 SLB accident, no reanalysis was performed. Thus, it is concluded that the reference cycle analysis is bounding for Cycle 15 operation.

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Page 59 of 67

rc . . i 7.00 1

6.00 N

CYCLE 15 HFP , CYCLE 8 HFP 5'00

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CYCLE 1 5 HZP - t. +

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200 250 300 350 400 450 500 550 600 Steam Line Break Accident Omaha Public Power District Figure l Reactivity vs. Moderator Temperature Fort Calhoun Station Unit No.1 7.3.2- 1 ,

Page 60 of 67

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7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) 7.3.3 Seized Rotor Event The Seized Rotor event was reviewed for Cycle 15 to demonstrate that only a small fraction of fuel pins are predicted to fail during this event. The analysis showed that Cycle 15 is bounded by the reference cycle (Cycle 9) analysis because an Fn of 1.85 was assumed in the Cycle 9 analysis and the Cycle 15 proposed COLR FI value of 1,77 remains conservative with respect to the Cycle 9 analysis value.

Therefore, the total number of pins predicted to fail 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. ,

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Page 61 of 67 I I

c, 8.0 ECCS PERFORMANCE ANALYSIS

' Both Cycle 15 Large and Small Break Loss of Coolant accident analyses were reviewed by Westinghouse based on the methodology discussed 'in Reference 1.

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.4 Page 62 of 67

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s 9.0 STARTUP TESTING The startup testing program proposed for Cycle 15 is identical to that used in Cycle 14. It is also the same as the program outlined 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 measurements will be substituted for the full core symmetry checks.

The CEA exchange technique is a method for measuring rod worths which is faster and produces less waste than the typical boration/ dilution method. The startup testing method used in Cycles 11,12,13 and 14 employed the CEA exchange technique exclusively. Results from the CEA exchange technique were within the acceptance and review criteria 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 azirnuthal power tilts during reload core physics testing. The pseudo-ejection rod measurement 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 otl,er symmetric CEA's within Bank 4 to measure od worths.

The acceptance and review criteria for these tests are:

Test Acceptance Criteria Review Criteria CEA Group Worths 15% of predicted 15% of predicted Using CEA Exchange Technique Pseudo-ejection None 1.5c deviation from rod worth group average measurement Low Power CECOR Technical Specifica- Azimuthal tilt less than maps tion limits of Fl, 20%.

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

Page 63 of 67

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9.0 STARTUP TESTING The startup testing program proposed for Cycle 15 is identical to that used in Cycle 14. It is also the same as the program outlined in the Cycle 6 Reload Application, with two exceptions. First, a CEA exchange technique (Reference 1) for zero power rod wodh measurements will be performed in accordance with Reference 2, replacing the ,

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

The CEA exchange technique is a method for measuring rod worths which is faster and produces less waste than the typical boration/ dilution method. The startup testing method used in Cycles 11,12,13 and 14 employed the CEA exchange technique exclusively. Results from the CEA exchange technique were within the acceptance and review criteria 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 the lead bank (Bank 4) into the core, borating a Bank 4 CEA out, and then exchanging (rod swap) the CEA against other symmetric CEAs within Bank 4 to measure rod worths.

The acceptance and review criteria for these tests are:

Test Acceptance Criteria Review Criteria CEA Group Wodhs 15% of predicted 15% of predicted Using CEA Exchange Technique Pseudo-ejection None 1.5c deviation from rod worth group average measurement Low Power CECOR Technical Specifica- Azimuthal tilt less than maps tion limits of FI, 20%.

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

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Page 63 of 67 1

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10.0 REFERENCES

References (Chapters 1-5)

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

OPPD-NA-8301-P, Revision 05, January 1993.

2. " Omaha Public Power District Reload Core Analysis Methodology - Neutronics  !

Design Methods and Verification", OPPD-NA-8302-P, Revision 03, January 1993.

3. " Omaha Public Power District Reload Core Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-P, Revision 04,  !

January 1993.

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

January 1987.

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

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10.0 REFERENCES

(Continued)

References (Chapter 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 Axial 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 Amendment No. 92 to Facility Operating License No. DPR-40 for Omaha Public Power District, Fort Calhoun Station, Unit No.1, Docket No. 50-285, November 29,1985.

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

OPPD-NA-8301 -P, Revision 05, January 1993.

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

7. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment 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.

8. Safety Evaluation Report on CENPD-207-P-A, "CE Critical Heat Flux: Part 2 Non-Uniform Axial Power Distribution", letter, Cecil Thom'as (NRC) to A. E.

Scherer (Combustion Engineering), November 2,1984.  ;

1

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10.0 REFERENCES

(Continued) ,

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References (Chapter 7) i

1. " Amendment No.117 to Operating License DPR-40, Cycle 12 License  !

Application", Docket No. 50-285, December 14,1988. .l

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

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

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

4. " Statistical Combination of Uncertainties Methodology, Part 3: Departure from Nucleate Boiling and Linear 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 Core Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-R Revision 04, January 1993.

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

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

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

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

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

12. " Omaha Public Power District Reload Core Analysis Methodology.- Neutronics Design Methods and Verification", OPPD-NA-8302-R Revision 03, January 1993.

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

dated November 8,1989.

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

dated July 31,1991.

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r, .

10.0 REFERENCES

(Continued) 1 References (Chapter 8)

1. " Omaha Public Power District Reload Core Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-R Revision 04, January 1993.

2. Letter LiC-91-247R, W. G. Gates (OPPD) to Document Control Desk (NRC),'

dated September 30,1991.

References (Chapter 9)

1. " Control Rod Group Exchange Technique," CEN-319, November 1985.

2. " Acceptance for Referencing ci Licensing Topical Report CEN-319 - Control Rod Group Exchange Technique," letter, Dennis M. Crutchfield (NRC) to Rik W.

Wells (Chairman - CE Owners Group), dated April 16,1986.

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