ML20094J116

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Application for Amend to License DPR-40,reflecting Necessary Changes for Cycle 14 Refueling Operation W/Attachments on Revised Inputs for Transient,T/H & Setpoint Analyses & Justification for Use of New Uncertainties
ML20094J116
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Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 03/06/1992
From: Gates W
OMAHA PUBLIC POWER DISTRICT
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NUDOCS 9203120078
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{{#Wiki_filter:_ - _ _ _ _ - _ - _ - - -- .. BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the datter of ) 

                                                                      )

Omaha Public Power District ) Docket No. 50-285 (Fort Calhoun Station ) Unit No. 1) ) APPLICATION FOR AMENDMENT OF OPERATING LICENSE cauant to Section 50.90 of the regulations of-the U. S. Nuclear R' ,ulatory Comn.ission ("the Commission"), Omaha ?ublic Power District, holder of Facility Operating License No. DPR-40, herewith requests that. Technical Specification 2.10 of the Technical Specifications set forth in Appendix A to that License be amended to reflect changes necessary for Cycle 14 operation. The proposed changes in Technical specifications and a Discussion, Juscification and Ho Significant Hazards Consideration Analysis, which demonstrates that the proposed changes do not involve significant hazards considerations, was appended as Attachments A and B respectively.of an Application of Amendment dated November 27, 1991. The attachod information provides corrected information for the Discussion of changes included in the Application dated November 27, 1991. The proposed changes in specificacions wvuld not authorize any change in the types or any. increase in the amounts of effluents or any change in the authorized power level of the facility. WHEREFORE, Applicant respectfully requests that Section 2.10 of Appendix A to Facility Operating License No. DPR-40 be amended in the form attached as Attachment A to the Application of Amendment dated November 27, 1991. (LIC-91-0320A) 9203120078 920306 PDR ADOCK 05000285 p- PDR

UNITED STATES OF AMERICA , NUCLEAR REGULATORY COMMISSION i In the Matter of ) 

                                                                             )

, Omaha Public Power Dictrict ) Docket No. 50-285 l (Fort Calhoun Station ) , Unit No, 1) ) AFFIDAYl1 , W. G. cates, being duly sworn, hereby deposes and says that he is.the Division Manager - Nuclear Operations of the Omaha Public Power District; that as such he . is duly authorized to sign and file with the Nuclear Regulatory Commission the attached information which provides corrected information for the Discussion of changes included in the Application of Amendment dated November 27,.1991-(LIC-91-0320A) concerning changes necessary for cycle 14 operations that he is familiar with the content thereof; and that the matters not forth therein are true and correct to the best of his knowledge, information, and iselief. JQ. Division Manager s Nuclear Operationo STATE OF NEBRASKA) 

                                      -)     as l       COUNTY OF DOUGLAS)
                                                                                                                                                                          ~

l Subscribed and sworn to before me, a. Notary Public in and for.t'3 Statn C; (. Nebraska on this 6 6 day of March, 1992.. [ IBMBE WARY Stat of Adrans ; 

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hotiry Publu h>EL My Casm. htL Febr. 27,1994 [ h r 

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A copy of this Application, including its attachmente, has been sub-mitted to the Director - Nebraska State Civision of Radiological Health, as requirnd by 10 CFR 50.91. OMAHA PUBLIC POWER DISTRICT By  ! * 

                                                                                                                        -/

Division Manager Nuclear Operations subscribed and sworn to before me t.his day of March, 1992. b i ( [ M)g/ f% 

                                                                 /  ,                JUDITH E w w.MAgitt A , , , , "-                                          w zum

(( NotaryPbl[c

i TABLE 1 

  ~

COMPARISON OF ORIGINAL AND REVISED INPUTS FOR THE CYCLE 14 TRANSIENT, T/H AND SETPOINT ANALYSES ITEM CYCLE 14 CYCLE 14 REVISED CHANGE CONSERVATIVE r

1. Minimum Available Scram Worth With Most Reactive Rod Stuck 100% Out, %Ap a). LOCA, LOFA. Loss of AC, Loss of Load.  ;

Seized Rotor, and CEA Withdrawal  ! HFP / BOC (PDIL) 6.4074 6.4711 Increase YES HFP / EOC (PDIL) 7.6272 7.6793 Increase YES HFP / BOC (ARO) 6.7244 6.7920 Increase YES HFP / EOC (ARO) 7.9573 8.0129 . Increase YES b) Loss of Feedwater Flow. Excess Load and Steam Generator Tube Ruoture ! HFP / BOC (PD!L) 5.7922 5.8545 increase YES HFP / ECC (PDIL) 6.3493 6.3656 Increase YES c) Steam Line Break Ever;t HFP / BOC (PDIL) 6.2885 6.3523 increass YES HFP / EOC (PD!L) _ 7.0243 7.0765 lacrease YES i HFP / BOC (ARG) 6.6055 ' 6.6732 increase YES MFP / EOC (ARO) 7.5545 7.4101 increase YES HZP / BOC (PDIL) 5.0596 5.1035 increase VES HZP / EOC (PDIL) 5.9833 5.9882 Increase YES HZP / BOC (ARO) 6.2280 62942 Increase YES HZP / ECC (ARO) 7.2823 7.3020 increase YES + 

                                                                                                                                                                     .b
      .       _      ___a                    - .                 ..-                  .             .,   --      . - . , . + - . .+_      ...      . _ _ . _ .-   -.

i TABLE 1 i (continued) COMPARiSOM OF ORIGINAL AND REVISED INPUTS FOR THE CYCLE 14 TRANSIENT, T/H AND SETPOINT ANALYSES ITEM CYCLE 14 CYCLE 14 REVI3ED CHANGE CONSERVATIVE

1. ' Minimum Availab!e Scram Worth With .

l Mast Reactive Rod Stuck 100% Out, %Ap (Continued) d) All HZP Events Except CEA Ejection  ; i ). HZP / SOC (PDIL) 5.0596 5.1035 - Increase YES HZP / EOC (PDIL) 5.9833 5.9082 increase YES 

                                                                                                                                                           -{

62942 HIP / BOC (AHO) 6.2280 increase ' YES ,l HZP/ EOC (ARO) 7.2823 7.3020 increase YES e) CEA Ejection - HFP / BOC (PDl'.) 5.7879 - NOT NECESSARYTO RECALCUMTE HFP / EOC (PDiL) _ 6.5936 THESE VALUES DUETO CEA  : 

..                     - HZP / BOC (PDIL) .                           4.3085                   EJECTION ANALYSIS CONSERVAriSMS                              ,

HZP / EOC (PDlL) l '4.0242 l f

1 e

g 3 4' _ _ _ __. . _ . . . _ .m. . . . ~ . - _- _ . , , . , . ..-

4 TABLE 1 (continued) l COMPARISON OF ORIGINAL AND REVISED INPUTS t '- FOR THE CYCLE 14 TRANSIENT, T/H AND SETPOINT ANALYSES l CYCLE 14 Ucense App!ication CYCLE 14 REVISED CHANGE CONSERVATIVE ITEM

2. Radial Peaking Factors a) P!anar Radial Peaking Factor (Fn) Decrease YES ARO 1.3494 1.8386 1.9916 1.9924 Increase NO Bank 4 2.0155- 2.0126 Decrease VES  :

Banks 4+3 l b) Integrated Radial Peaking Factor l (F,) 1.7831 1.7856 increase - NO' ARO 1.9053 1.9122 increase NO Bank 4 - 1.9229 1.9233 increase NO Banks 4+3 l 

                                                                                                          . . . . _ - .- j

TABLE 1 (continued) COMPARISON OF ORIGINAL AND REVISED INPUTS FOR THE CYCLE 14 TRANSIENT, T/H AND SETPOINT ANALYSES ITEM CYCLE 14 CYCLE 14 REVISED CHANGE CONSERVATIVE

3. CEA V!ithdrawal Data

, Maximum DifferentialWorth, Aplin. 3.5133E-04 3.6059E-04 increase NO

4. CEA Drop Data c) 3-D / HFP, BOC (Biases and Uncertainties included)

ARO . Distortion Factor 1.1915 1.1822 Decrease YES Rod Worth. %Ap 02947 02871 Decrease NO Bank 4 (PDll) . Distortion Factor 1.1884 1.1796 Decrease YES Rod Worth, %Ap 02940 02863 Decrease NO b) 3-D /100% Power,500 MWD /MTU (Biases and Uncertainties included) ARO Distortion Factor 1.1937 1.1818 Decrease YES ,i Rod Worth, %Ap N/A 02887 Bank 4 (PDIL) . Distortion F3ctor 1.1904 1.1812 Decrease YES Rod Worth, %Ap . N/A 02880 

      . c) 3-D / 62.6% Power, BOC (Biases and Uncertainties included)

ARO Distortion Factor 1.2330 12240 Decrease YES Rod Worth, %Ap 02628 02548 Decrease NO Bank 4 (PD!L) Distortion Factor 12243 N/A

  • Rod Worth, %Ap 02634 N/A d) 3-D / HFP EOC (Biases and Uncertainties included)

ARO Distortion Factor 1.1570 1.1495 Decrease YES Rod Worth, %Ap -0.3461 0.3361 Decrease NO Bank 4 (PDIL) vistortion Facter. 1.1549 N/A ' Rod Worth, */Ap 0.3424 N/A 

                                                                           --r-- - - -

h ATTACHMENT 1 1

2.0 OPERKilNG HISTORY OF CYCLE 13 Fort Calhoun Station is presently operating in its thirteenth fuel cycle utilizing Batches L, M, N and P fuel assemblies. Fort Calhoun Cycle 10 operation began when criticality was achieved on May 25,1990, and full power reached on June 18,1990. The reactor has operated up to the present time with the core reactivity, power distributions, and peaking factors having closely followed the calculated predictions. Actual Cycle 13 termination burnup of 15,248.82 MWD /MTU was achieved on February l 1 1992, and was used as a basis for the reanalysis of Cycle 14. Page 5 of 62

l 3.0 GENERAL DESCRIPTION (Continued) The fuel rod and poison rod locations in Batches M and N shimmed assemblies arc shown in F!gure 3-2. Figure 3-3 shows the fuel rod locations in Batches N and P unshimm9d assembt!es. 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 lowere the initial enrichment of U-235 in those fuel rods while maintaining an assembly average initial enrichment sufficient to achieve the Cycle 14 design exposure. Figure 3-5 shows the fuel rod locations for the Batch R1 natural uranium assemblies. Fiqures 3-6 through 3-9 provide a diagram of each type of fresh assembly which conte.ns IFBA rode. The average initial Enrichment cf the 52 fresh Batch R assemblies is 3.57 w/o U-235, a reduction of 0.09 w/a from Cycle 13. Excluding the four fresh natural uranium assemblies, tM es je initial enrichment is 3.81 w/o U-235. For the second consecutive cych a ;ci assembly zone loading technique is used to lower the radial power peaking fa c; M' in Batches R2 through R7. Batch R2 through R5 assemblies have fuel rods at both *.v w/o enriched U-235 and 3.5 w/o enriched U-235, while Batch R6 and R7 assemblies have fuel rods at both 3.75 w/o enriched U-235 and 3.25 w/o enriched U-235. Figure 3- 10 shows the beginning of Cycle 14 assembly burnup distribution for a Cycle 13 termination burnup of 15,250 MWD /MTU The initial enrichment of each fuel g assembly is also shown in Figure 3- 10. Figure 3-11 shows the projected end of Cycle 14 assembly burnup distribution. The end of Cycle 14 core average exposure will be approximately 28,547 MWD /MTU. 5 Page 7 of 62 l _______o

l TABLE 3-1 FORT CALHOUN UNIT NO.1 CYCLE 14 CORE LOADING Initial BOC EOC Poison IFBA Poison Assembly Number of , Burnup Avg. Burnup* Rods per Rods per Loading Designation Assemblies ' AvpWD/pg MWD /uru Assembly Assembly gm 82 p/n, M/ 1 31,408 46,196 8 - 0.024 l N 20 28,803 38,932 0 - 0 l N/ 20 32,229 3e 574 8 - 0.020 l P 8 13,702 30,342 0 - 0 l P/ 32 19,298 34,464 8 - 0.027 l R1 4 0 4,421 - 0 0.003 l R2 16 0 13,868 - 28 0.003 l R3 4 0 19,962 - 48 0.003 l R4 8 0 19,885 - 64 0.003 l R5 12 0 20,891 - 84 0.003 l R6 4 0 20,598 - 84 0.003 l R7 4 0 19,112 - 64 0.003 l Assumes EOC14=14,000 MWD /mu Page 8 of 62 l

2 J AA - Assemblylocation i BB - Fuel Typo C.CC - Initial Enrichment (w/o U-235) i DD,DDD - Assembly Average Exposure (MWD /MTU) l l 2 1' N N/ - 3,70 3.70 24,691 30,556 3 4 5 6 7 , N/ - R2 P/ R2 P/-- < 3.70 3.85 3.59 3.85 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- , 20 21 22 23 24 25 ' P/ R4 P/ P/-- N/ R6 26 3.59 3.85 '3.59 3.59 3.70 3.60 16,964- 0 21,006 15,148- 30,506 N 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 '- N/ 0' 31,077 0 30,880 0 21,059 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 i Cycle 14 BOC Initial Enrichment  : Omaha Public Power District J Figure and Assembly Average Exposure Fort Calhoun Station Unit No.1 3-10 Page 18 of 62.

s 2 AA - Assembly Location BB - Fuel Type C.CC - Initial Enrichment DD DDD -- Assembly Average (w/o(MWD Exposure U-235) /MTU) 

                                                                                                  'N3.70
  • N/

3.70 28,583-- 33,931 4 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 i 8 9 10 11 12 13 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 3.94 3.85 3.59 3.85 3.59 4,371 29,441 20,891 38,317- 20,673- 38,469 20 23 21 22 24 25 P/ R4 P/ P/ N/ - R6 26 3.59 3.85 3.59 3.59 3,70 3.60 N 27,565 18.788 38,166 -32,754 45,878 20,383 3.70 27 28 29 30 31 32 29,286 R2 N R5 N R4 P/- 33 3.85 3//0 3.85 . 3.70 - 3.85 3.59 15,354 45,250 20,652. 45,896 20,620 37,929 N/ 3.70 34 35 36 37 38 39 35,955 R3 P/ P/ R6 P/-. M/ 3.59 3.85- 3.59 3.60 3.59 3.80 33,981 20,084 38,069 20,456 38,007 45,607 , i l

Cycle 14 EOC Initial Enrichment Omaha Public Power District Figure L and Assembly Average Exposure Fort Calhoun Station Unit No.1 3 - 11 l Page 1.9 of 62 l _ .

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 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 vessellimiting welds. This extreme low radialleakage fuelloading 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 14 have been reduced by lowering the enrichment of the fuel 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 design. 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 burned 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 of 15,250 MWD /MTU and limiting values established for the safety analysis. 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 cisplayed iri 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 are the most limiting conditions used in the determination of available shutdown margin for compliance  !

with the Technical Specifications. The minimum available shutdown I margin is 1.06%Ap with respect to the Technical Specification limit of l

l Page 21 of 62

5.0 NUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.1 Fuel Management (Continued) 4.0%Ap. Table 5-2 presents a summary of CEA shutdewn worths and reactivity allowances for Cycle 14. The cycle 14 CEA worth values, used in the calculation of minimum scram worth, exceed the minimum value required by Technical Specifications and thus provide an adequate shutdown margin. 5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC14, MOC14, and EOC14, respectively, and are based upon the Cycle 13 late window burnup timepoint. These radial power densities are assembly averages representative of the entire core length. The high burnup end of the Cycle 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 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 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 l peaking to be observed during operation. As previously indicated, Figures 3-5 and 3-6 show the integrated assembly burnup values at 0 and 14,000 MWD /MTU for Cycle 14. l 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 14 during normal base load, all rods out operation at full power is 2.1069, including uncertainty allowances. l ; l l 1 l Page 22 of 62 l l

TABLE 5-2 FORT CAUiOUN UNIT NO.1, CYCLE 14 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES ** FOR HOT ZERO POWER BOC, HZP EOC, HZP l (%Ap) (%Ap) g

1. Worth of all CEAs inserted 7.52 8.86 E
2. Stuck CEA Allowance 1.17 1.43 E
3. Worth of all CEAs Less Worth of Most Reactive CEA Stuck Out 6.35 7.43
4. Power Dependent insertion Limit CEA Worth 1,19 1.33
5. Calculated Scram Worth 5.16 6.10 I
6. Physics Uncertainty plus Bias 0.10* 0.12* I
7. Net Available Scram Worth 5.06 5.98 E
8. Technical Specification E Shutdown Margin 4.00 4.00 5
9. Margin in Excess of Technical l Specification Shutdown Margin 1.06 1.98 g 1.96% of calculated scram worth from revised ABB-CE methodology biases and E uncertainties.
 ** These values are the same values as the original analysis, prior to detection and correction of the hafnium- related cross-section error. The results remain conservative with respect to the corrected scram worths, i.e. the above values are less than the revised values.

Page 25 of 62

AA - Assembly Location D.DDDD - Assembly Relative Power Density b "- C.CCC - Maximum 1-Pin Peak Assembly , 1 1 2 0.2751 0.2347 l 3 4 b G 7 0.3721 0.0110 0.8667 1.0990 0.0673 8 9 10 11 12 13

0.2018 1,0680 1.3302 1.3466 1.0139 1.4143  !

14 15 16 17 18 10 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 1,0004 1.3210 0.3264 ' 27 $8 29 30 31 32 1.1522 1,0163 1.3734 1.0589 1.4306 1.2305 33 0.3937 34 35 36 37 38 39 0.9043 1.4475 1.2461 1.3249 1.2349 1.1051 1 Maximurn 1 -Pin Peak at 23% Core Height F Cycle 14 Assembly Power Distribution Omaha Public Power District Figure O MWD /MTU, HFP, Equilibrium Xcnon Fort Calhoun Station Unit No.1 5-1 Page 27 of b2

AA - Assembly Location B.BBBB - Assembly Relativo Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.2661 0.2308 3 4 5 6 7 0.3440 0.8433 0.8142 1.0429 0.8104 a 8 9 10 11 12 13 0.2007 0.0873 1.2004 1.3964 0.9951 1.4327 14 15 16 17 18 19 0.3106 1.1085 1.5300 1.229G 1,5334 1.2628 20 21 22 23 24 25 26 

                                      ~                                                       ~

27 28 20 30 31 32 1.0801 0.9987 1.5319 ' 1.0716 1.5103 1.2054 33 1,7370 34 35 3G 37 38 39 0.8815 1.4567 1.2730 i.0309 1.5311 1.2095 Maximum 1-Pin Peak at 23% Core Height C Omaha Public Power District Figure 

      ! ycle 7,00014 MWD Assembly
                      /MTU, HFP, Power Eq: XenonDistribution Fort Calhoun Station Unit No.1                                            5-2           .

Page 28 of 62 i _ __ _ _ o 

 - -                            - ~. -                                         - . -       - . - . - . . - . -               - - - . . - - -                             . - . . - . - - - . . -

i i t i r 3 AA - Assembly Location B.BBBB - Assembly Relativo Power Density C.CCC - Maximum 1-Pin Peak Assembly - i 1 2 0.3175. 0.2775 . 3 4 5. 6 7 0.3963 0.9056 0.8645 1.1034 0.8602 l i 8 9 10 11 12 13 0.2426 1.0402 1.1008 1.3758 0.9966 1.0986 i i 14 15 16 17 18 19 l 0.4078 1.1243 1,4778 1.1698 1.4593 1.1899 1.6550  ! 20 21 22 23 24 25 l 0.8223 1.3852 1.1702 1.1710 1.0403 1.4233 . 20 0.3851 -- + 27 28 29 30 31 32 , 1.1709 -1.0130 1.4674 1.0194 1.4047 1.1143  ! 33 . 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. , i 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

4 AA - Assembly Location B.l3BBB - As:,ombly Relativo Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.2949 0.2571 3 4 5 6- 7 l 0.2572 0.8059 0.8772 1.1819 0.9490 8 pys ff 10 11 12 13 

                                                                                                                                                                                              ]

0.1284 4 ~ 1.1595 1.3600 1.0870 1.5387 ' hkl$ i 14 15 18 17- 18 19-0.2191 1,0341 1.3964 1.3206 1.4719 1.3280 20 21 22 23 24 25 0.7986 1.3397 1.3099 1.3957 1.1409 1.3719 20 0.3633 27 28 29 30 31 32  ; , 33 , 0,4444 34 35 38 37 38  % 4 1.0791 1.6002- 1.3582' 1.3824 1,1438 L g 3. ' 1.7810 b Maximum 1-Pin Peak at 20% Core Height g~ = - Bank 4 Locations  ; Cycle 14 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 30 of 62 . E____.___.___ -. _ _ . _ . _ . _ . . ... . _ _ . ~ . . _ _ . . _ _ . - _ . _ _ _ . _ . - . ,

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

0.0404 0.3033 3 4 5 6 7 O.2691 0.7999 0.8769 0.1847 0.9387 8 [Mg 10 11 12 13 O.1557 1.0364 1.3930 1.0665 1.5173 14 15 16 17 18 19 0.3647 0.9855 1.4196 1.2066 1.5563 1.2808 r

20 21 22 23 24 25 e 28 O.4241 27 28 29 30 31 32  : 1.2807 1.0974 1.5749 1.0737 1.0434 1.0474 33 1.8000 0.5019 34 35 36 37 38' $- $ 1.0501- 1.5686- 1.3088 1 4985 1.0519 (( sq g  ; i, i. Maximum 1-Pin Peak at 17% Core Height 7 

                                                                                                 %1    - Bank 4 Locations l

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

G.0 THERMAL-HYDRAULIC DESIGN G.1 D.NBR ANALYSIS Steady stato DNBR analyses of Cyclo 14 at the rated power of 1500 MWt have been performed using the TORC computer codo described in Referenco 1 and the CE- 1 critical heat flux correlation doccribed in Ref9tenco 2. The CETOP-D computer code described in Referenco 3 was used in the setpoint analycis, but was replaced by the TORC code for DNBR analyses. The DNBR analysis applications and methods did not chango from previous cyclos, with the exception that the TORC computer codo was used to calculato the minimum DNBR rather than the CETOP-D computer code. Both codes are approved for use with the OPPD methods. This is difIeront from the combination that was used in the Cyclo 8 through Cycle 13 Fort Calhoun reload analysos (References 4 through 9). The reload methodology for Cyclo 14 can be found in Reforence 10. Tabic 0- 1 contains a list of portinent thermal-hydraulic parameters used in both safety analyses and for generating reactor protective system setpoint information. The calculational factors (engineering heat flux factor, enginocring Iactor on hot channel heat input, rod pitch and clad diameter factor) listorlin Table 6-1 have been combined statistically with other uncertainty factors at the 95/95 confidence / probability level (Referenco 11) to define the design limit on CE-1 minimum DNBR. 6.2 FUEL ROD BOWING The fuel rod bow penalty accounts for the adverso impact on MDNBR of random variations in spacing between fuel rods. Tho penally at 45,000 MWD /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 uncertainties (Referenco 11). The design basis for the amount of fuel rod bow allowed in the Westinghouse fuel and for the CE fuel design is the same. Westinghouse has identified in the mechanical fuel design report that the amount of dellection does not requiro a DNB penalty to be applied under Weslinghouse analysis requirements. Thus, the CE DNB penalty was applied to the l West!nghouse fuel to ensure that the OPPD statistical combination of uncertainties woro r.till valid and that conservative input assumptions woro used l In the analysis. Page 32 of 62 i

I 1 l TABLE 6-1 FORT CALHOUN UNIT NO.1, CYCLE 14 THERMAL HYDHAULIC PARAMETERS AT FULL POWEH Unit Cyclo M* Total Heat Output (Coro Only) MWt 1500 10' BTU /hr 5110 Fraction of Heat Generated in Fuel Rod 0.975 Primary System Pressuro Nominal psla 2100 Minimum in Steady Stato psia 2075 Maximum in Steady Stato psia 2150 Infot Temperaturo (Maximum) 'F 545 g Total Reactor Coolant Flow ppm 202,500 (Steady State) 10 lbm/hr 76.32 (Through the Core) 10* lbm/hr 73.06 Hydraulic Diamator (Nominal Channel) ft 1044 Averago Mass Velocity 10' ibm /hr-It" 2.226 Core Averago Heat Flux (Accounts for Heat Generated BTU /hr-ft' 181281 in Fuel Rod) Total Heat Transfer Surface Area (t' 20,241** Averago Coro Erdhalpy Rise BTUllbm 72.0 Average Unear Heat Rato kW/ft 6.01 *

  • Engineering Heat Flux Factor 1.03 * *
  • Engincoting Factor on Hot Channel Heat input 1.03 * *
  • Rod Pitch and Bow 1.005 * *
  • Fuel Densification Factor (Axial) 1 002
  • Design inlet temperature and nominal pr'. mary system pressure were used to l

calculato these paramotors. 

            *
  • Based on Cycle 14 specific value of 4.?4 fuel displacing shims.
        *** Tnose factors woro combined statistically (Reference 8) with other uncertainty f actors at 95/95 confidence / probability lovcl to define a design limit on CE-1 l                minimum DNBR.

! Page 33 of 62

TABLE 7-1 FORT CALHOUN UNIT NO.1, CYCLE 14 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS 7.1 Anticipated OperationalOccurrences for whichintervention of the RPSis nocessary to provent exceeding acceptablo limits: 7.1.1 Reactor Coolant System Depressurization Reanalyzed 7.1.2 Loss of Load Not Roanalyzed5 7.1.3 Loss of Feodwater Flow Not Roanalyzed5 7.1.4 Excess Heat Removal due to Feodwater Malfunction Not Roanalyzed5 7.1.5 Startup of an inactivo Reactor Coolant Pump Not Roanalyzedi 7.2 Anticipated Operational Occurrences for which sufficient initlnl steady stato thermal margin, maintained by the LCOs, is necessary to prevent excooding the acceptable limits: 7.2.1 Excess Load Roanalyzeda 7.2.2 Sequential CEA Group Withdrawal Reanalyzed 2 7.2.3 Loss of Coolant Flow Roana!yzed3 3 7.2.4 CEA Drop Roanalyzed 7.2.5 Boron Dilution Reviewed 7.2.6 Transients Resulting frorn the Malfunction of One Steam Generator Not Roanalyzed4 7.3 Postulated Accidents 7.3.1 CEA Ejection Roanalyzed 7.3.2 Stcam Line Break Reviewed 5 7.3.3 Solzod Rotor Roanalyzed5 7.3.4 Steam Generator Tube Rupturo Not Reanalyzed 1 Technical Specifications preclude this event during cgoration. 2 Requires High Power and Variablo High Power Trip. 3 Requires Low Flow Trip. 4 Requires trip on high differential steam generator pressure. 5 Event bounded by referenco cycle analysis. A negativo dolormination utilizing the 10 CFR 50.59 critoria was mado for this event-Page 35 of 62

i TABLE 7-2 i FORT CALHOUN UNIT NO.1, CYCLE 14 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND OTM (CENTERLINE TO MELT) DESIGN LIMITS p'hysics Parameteis Units Cycle 13 Values Cycle 14 Values Radial Peaking Factors For DNB Margin Analyses (F') Untodded Region 1.70* 1.79* Bank 4 Inserted 1.73* 1.92* For Planar Radial Component (Fk ) of 3- D Peak (CTM Limit Analyses) Unrodded Region 1.75* 1.85* Bank 4 Inserted 1.77* 2.0* Maxirnum 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 Umiting EOC Zero Power SLB) %Ap - 4.0 -4.0 The DNBR analyses utilized the methods discussed !n Section 6.1 of this report. The procedures used in the Statistical Combination of Uncertainties (SCU) as they portcin to DNB and CTM limits are detailed in References 2-5. I Page 36 of 62

TABLE 7-3 FORT CALHOUN UNIT NO.1 DESIGN BASIS EVENTS REANALYZED FOR CYCLE 14 Reason for Acceptance Summary Event _ Reanalysis Critoria of Results Sequential CEA Calculate cycle specific Minimum DNBR 2 MDNBR =1,71 l J 1.18 using the CE-1 Group Withdrawal ROPM values PLHOR< 22 kW/ft correlation. Transient PLHGR g 22 kW/ft. CEA Drop incorporated bounding Minimum DNBA 2 MDNBR = 1.30 j input values 1.18 using CE-1 PLHGR < 22 kW/ft l correlation. Transient. PLHGR $ 22 kW/tt Excess Load Reclassified as a ROPM ovent Minianum DNBR 2 MDNBR = 1.31 (methodology chango) 1.10 using CE-1 PLHGR < 22 kW/ft correlation. Transient PLHGR $ 22 kW/ft RCS Depressurization To provido a conservativo Pbias Pblas vatuo s the Pblas = 30 psia input for the TM/LP due to the previous cycle's limiting Excess Load methodology value (from Excess Load chango and RCS Deproscurization) Loss of Coolant flow To provide for inctuased inst- Minimum DNBR 2 MDNBR = 1.42 rument uncertainty in the RPS 1.18 using CE-1 PLiiGR < 22 kW/ft low flow trip circuit. This would correlation. Trans ent. reduce the trip sotpoint to 90% of Pl.HGR $ 22 kW/ft full flow conditions. l l i l Page 38 of 62 1

7.0 @NSIEN f ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) 7.2.2 CEA Withdrawal Everu The CEA Withdrawril (CEA V) event was reanalyzed for Cycle 14 to determine the initia) margins that must be mr.intained by the Limiting Conditions for Ope;ations (LCOs) such that the DNBR and 'uol contorlino to melt (CTM) design limits will not be exceeded in conjunction with the RPS (Variablo High Pnwer, liigh Pressurizer Pressure, or Axial Power Distribution Trips). The methodology contained in Referenco 6 was employed in analyzing the CEAW event. This event is classiflod as ono for which the acceptablo DN3R and CTM limits are not violated by virtue of maintenance of sufficient initial steady stato thermal margin providad by the DNBR and Linear Heat Rate (LHR) related LCOs. For the HFP CEAW DNBR analysis, a f 1TC valuo ldentical to that utilized in Referenco 8 and a gap thermal conductivity consistent with the assumption of Referenco 6 were used in conjunction with a variable reactivity insertion rato The HFP case for Cyc'o 14 is considered to rnoot the 10 CFR 60.59 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 negativo 10 CFR 50.59 determination was made for Cycle 14, the conclusions for Cycle 12 remain valid and applicable to Cycle 14. l Tho zero power case was analyzed to demonstrate that acceptablo DNBR and centerlino inett limits are not excocded. For the zero power caso, a reactor trip, initiated by the Variablo High Power Trip at 30% ( 20% plus [ 10% uncertainty of rated thermal power) was assumod 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 referenco cycle. The zero power caso initiated at tho limiting conditions of operation results in a a minirnum CE-1 DNBR of 5.44 which is less than the Cycle 12 valuo l of 6.99, but stillfarin excess of the minimum 1.18 DNBR limit. The analysis shows that the fuel to conterline melt temperatures are well below those corresponding to the accepta ble fuel to cen terline melt limit. Tho key input parameters used for the zero power caso are presented in Table 7.2.2- 1. l Page 43 of 02

TABLE 7.2.2 1 FORT CALHOUN UNIT NO.1. CYCLE 14 KEY PARAMETERS ASSUMED IN THE HZP CEA WITHDRAWAL ANALYSIS Paramotor Units Cycle 12 Cyclo 14 InitJa! Core Power Level MWt 1 1* Coro Inlet Coolant Temperature F 532 532* Pressurizer Pressure psia 2050 2075* Moderator Temperature Coefficient x10-4 Ap/ F + 0.5 + 0.5 Doppler Coefficient Multiplier 0 85 

                                                               .                 0.85 CEA Worth at Trip                         %Ap                  5.28              5.048          l Reactivity Insertion Rate Range                           x 10-4 Ap/sec           0 to 1.0           0 to 2.7 CEA Group Withdiawal Rato                                     in/ min                46                 46 Holding Coil Delay Time                    sec                 0.5                 0.5 Tho DNBR calculations used the methods discussed in Section 0.1 of this docurnent and detailed in References 2 through 5. The effects of uncertairities 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 OPERATIO_NAL OCCURRENCES (CATEGORY 2) (Continued) , 7.2.3 Loss of Coolant Flow Event The Loss of Coolant Flow event was roanalyzed for Cyclo 14 to determino l the minimum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) such that in conjunction with the RPS low flow trip, the DNBR limit will not be exceeded. The ovent was analyzed parametrically in initial axlal shape and rod configuration using the methods described in Referenco 6 (which utilizes the statistical combination of uncertaintles in the DNBR analycis as described in Appendix C of References 4 and 5). The 4-Pump Loss of Coolant Flow produces a rapid approach to the DNBR limit due to the rapid decreaso in the coro coolant flow. Protection against excooding the DNBR iimit for this transient is provided by the initial i steady stato thermal mar 0ln which is maintained by adhering tho the , LCOs on DNBR margin and by the response of the RPS which provides an automatic reactor trip on low reactor coolant flow as measured by tho , steam generator differential pressure transmitters. The flow coastdown is generated oy tho CESEC-Ill (References 9 and 10) which utilizes implicit modeling of the reactor coolan' pumps. Tablo 7.2.3-1 lists the key transient parameters used in the Cycle 14 analysis and compares them to the reference cyclo (Cycle 12) values. The low flow trip setpoint is reached at 3.66 seconds and the scram rods start dropping into the core 1.15 seconds lator. A minimum CE-1 DNBR of 1.422 is reached at 5.5 seconds, it may be concluded that for Cyclo 14 the Loss of Flow event, when initiated from the LCOs, in conjunction with the Low Flow Trip, will not exceed the minimum DNBR design limit. Page 46 of 6P.

l l TABLE 7.2.3-1 l FORT CALHOUN UNIT NO.1, CYCLE 14 1 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Cyclo 12 Cyc!c 14 Initial Coro Power Level MWt 1500* 1500* Initial Core Inlet Coolant Temperature 'F 545* 545* Initial RCS Flow Rato 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 - 0.50 -0.72 LFT Analysis Setpoint  % of initial flow 03 00 g LFT Responso 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 (includirIO Holding Coll Delay Total Unrodded Radial Peaking Factor (FI) 1.80 1.79 l The uncertainties on these parameters woro combined statistically rather than deterministically. The values listed represent the bounds included in the statistical combination. Page 47 of 62

TABLE 7.2.4- 1 FORT CALHOUN UNIT NO.1, CYCLE 14 KEY PARAMETERS ASSUMED IN THE HFP CEA DROP ANALYSIS Parameter Units Cycle 11 Cycle 14 Initial Coro Power Level MWt 1500* 1500* Core Inlet Coolant Temperature 'F 543* 545* Pressurizer Pressure psia 2075* 2075* Core Mass Flow Rate gpm 202,500* 100,000* Moderator Temocrature Coefficient x10-4 Ap/* F ~ 2.7 - 3. 0 Doppler Coefficient Multiplier 1.15 1.40 CEA Insertion at Maximum Allowed Power Wlnscrilon of Bank 4 25 25 Dropped CEA Worth Unrodded, %Ap -0.2337 -0.2887 l PDIL, %Ap - 0.2295 -0.2880 g Maxlinum Allowed Power Shape Index at Negative Extreme of LCO Band - 0.18 -0.18 Radial Peaking Distortion Factor Unrodded Region 1.1568 1.1818 Bai,k 4 Inscried 1.1508 1.1812

  • The DNBR calculations used the methods discussed in Sect;on 0.1 of this document and deteiled in References 2 throu h 5. The effects of uncertainties on these parameters were accounted for statisticall in the DNBR and CTM calculations.

Page 49 of 62 l

4 TABLE 7.2.4-2 FORT CALHOUN UNIT NO 1, CYCLE 14 SFOUENCE OF EVEN1S FOR FULL LENGTH CEA DROP Time (sec) Event Setpoint or Value  ; l 0.0 CEA Begins to Drop into Coto --- i 1.0 CEA Reaches Fully inserted Position 100% insertion 1.14 Core Power Level Reaches a Minimum 64.4% of 1500 MWt l and Degins to Return to Power Due to Reactivity Feedbacks 74.9 Core Inlet Temperaturo Reaches a 538.85'F l Minirnum Value , 199.5 RCS Pressure Reaches a Minimum 1998.24 psia l Value 200 0 Core F'ower Returns to its Maximum 94.98% of 1500 MWt l Value 200.0 Minimuro DNBR is Reached 1.379 (CE-1 Correlation) l l Page 50 of 62

TABLE 7.2.5- 1 FORT CAUlOUN UNIT NO.1 CYCLE 14 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Parameters Cycle 13 Values Cyclo 14 Values Critical Boron Concentration, ppm (ARO, No Xenon) Modo Hot Standby 1602 1292 Hot Shutdowr, 1662 1292 Cold Shutdown - Normal RCS Volume 1457 1204 Cold Shutdown - Minimum RCS Volumo* 1270 1204 Refueling 1454 1180 Inverso Baron Worth, ppm /%Ap Modo 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 Assum9d, %Ap Modo 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 1700*** l Shutdown Groups A and B out, all Regulating Groups inserted except most reactive rod sinck out. 

           ** Includes a 5.0%ip shutdown margin.
           *** Praposed Cycle 14 COLR value.                                                         E Page 52 of 62

I 7,0 TRANSIENT ANALYSIS (Continued) 7.3 POSTUUgED ACCIDENTS (Continued) 7.3.3 Soized Rotor Event The Seized Rotor event was reanalyzed for Cyclo 14 to demonstrato that l orily a small fraction of fuel pins are predicted to fall during this event. Tho ar alysis showed that Cyclo 14 is bounded by the reference cycle (Cyclo 0)  ; or alysis b9cause an Fn of 1.85 was assurned in the Cyclo 9 analysis and l th) Cyclo 14 Technical Specification of 1,79 rernains conservative with l . rc spect to the F; valuo used in the Cyclo 9 analysis. Tnerefore, the total number of pins predicted to fail will continue to be less i than 1 % of all of tho fuel pins in tho coro. Based on this result, tho resultant j sde boundary dose would be well within the limits of 10 CFR 100. I r Page 56 of 6'2

m' As-.___m_Je u44__Aa.4maaAerA.M.4--r s - u c_ _4.AA_'-A.oam_m.xM2;w, 44.w.. p.A ,.h.- = e _sw a 4 . wa4. ._ A eAJ.4-m.em_44<C,e.mm.g.3m4 , m.,. e 4 4.sp- .4a- -. & d,. I i i ATTACHMENT 2 I f P i I L L i l l 

                                                                                            =

I l 2 

                                                                                                                                               '-waq .y,,,,

ATTACHMENT 3 The purpose of this attachment is to provide OPPD's justification for the use of the new calculational uncertaintlos for peaking factors derived from the ROCS-NEM methods. Uso of upgraded reactor physics codos necessitates the use of uncertainties and biases consistent with application of the now inuthods. It will be shown that poaking factors calculated using the now methods, with blases and uncertaintios (consistent with the now methods) appliod, ato comparable to those obtained using the former methods (with the former biases and uncertainties included). The NRC-approved referenco cyclo (i.e. Cyclo 13) reload submittal utilized the Higher Order Differenco (HOD) method which is described in Referenco 1. For the Cycle 14 submittal, the Nodal Expansion Method (NEM), which is also described in Reference 1, wan linpismonted to increase the calculational accuracy of the nuclear design codos. Opocific changes incorporating the now methods include:

1. Implomontotion of NEM into the ROCS codo;
2. Improvements in accountability of anisotropic scattoring and higher order interface current angular distributions in the DIT codo;
3. Introduction of assembly discontinuity factors betwoon the ROCS and DIT codos;
4. Updato of biases and uncertaintios appilod to calculated parameters.

The revised biasos and uncertaintios associated with the application of the NEM are described in Referenco 2, introduction of tho lmproved methods required the ro-ovaluation of the biases and uncertaintles. The ABB-Combustion Engineering data base used to establ!sh the biases and uncertaintios was expanded to reflect recent roload cyclos with low leakage and high bumup fuel management, The data base was derived from tho following sources which includes Fort Calhoun Station: Plant Cyclo Palo Vordo 1 2 Palo Verdo 2 2 Palo Verdo 3 2 Palo Verdo 1 3 Palo Vordo 2 3 Calvert Cliffs 1 10 Calvert Cliffs 2 8 Fort Calhoun 13 Total Cycles: 8 in addition, Calvert Cliffs 2, Cyclo 9 data was added later and found to be consistent with the above data base. For justification of the application of the NEM biases and uncertainties to the unrodded planar and integrated radial peaking factors, the uncertainty plus bias terms (1.0. upper tolerance limits) appropriato for uso are found in Reference 2, ,. Tablo D (Items D-S and D-6), page D-1. The NEM upper tolerance limits for pin 1 I _ _ - _ _ _ _ _ _ _ _ . l

ATTACHMENT a peaking factors Fxy and Fr are 5.35% and 4.00%, respectively. Using the former HOD method (Reference 1), upper toleranco limits for pin peaking factors F,y and Fr aro 4.99% and 3.02%, respectively, in order for both NEM and HOD methods to produce similar pin peaking factors, the NEM upper toleranco limits must be larger in valuo than the HOD upper tolerance ilm!!s. Therefore, the pin peaking factor upper toleranco limits for NEM ato moro conservative than the pin peaking factor upper toleranco limits for HOD. To verify that use of the HOD method and the NEM method produces similar results, a Cycle 13 pin peaking factor model using NEM was generated and compared to the Cycle 13 HOD results. The results of this comparison, along with the upper toleranco limits for both methods, are presented below along with results from Cycle 14 using NEM: FORT CALHOUN UNIT NO.1 MAXIMUM PIN PEAKING FACTORS AND ALLOWANCES n Cycle 13 Cycle 13 Cycle 14 (HOD) INEMJ ,(NE,M)

1. Upper Tolerance Lirnits (%),

Fry 4.990) 5.35(2) 5.35(2) F, 3.020) 4.00(2) 4.00(2)

2. Calculated Maximum Peaking Factors, (from ROCS)

Fxy 1.585 1.58B 1.745 Fr 1.553 1.558 1.717

3. Final Maximum Peakina Factors, (Calculated + Upper Tolerance)

Fxy 1.664 1.673 1.839 Fr 1.600 1.621 1.786

0) HOD Upper Toleranco (Jmit from CENPD-266-P-A (2) NEM Upper Toleranco Limit from CE-CES-129, Revision 1-P The results show that the differences between the HOD and NEM methods for calculating the Cycle 13 Fxy and Fr maximum pin peaking factors to bo 0.54% and 1.31%, respectively, which are considered to be acceptably small, it can also be calculated from the above results that NEM produces slightly more conservativo maximum pin peaking factors than HOD.

In summary, the application of the revised upper tolerance limits to the pin peaking factors (as presented in Reference 2) are consideiod to be justified for application with the NEM method rather than the HOD method. Both methods are described in Reference 1. OPPD's propost'd application of NEM in Cycle 14 was consistent with the samo method used in the ud.'vation of the biases and uncertainties of Reference 2. 2 

                                                      ..a. _ . . . . . . .         .

ATTACHMENT 8

References:

1. "The ROCS & DIT Computer Codes for Nuclear Design", CENPD-206-P-A.

April 1983, -

2. " Physics Diasos and Uncertaintles", CE-CES-120. Revision 1-P Abgust 1991.

1 l s k t c 4 e i I h I y [' l 6 3 T" 

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