ML19340A149

Amend to BAW-1432,Oconee 3,Cycle 2 Reload Rept.

ML19340A149
Person / Time
Site:	Oconee
Issue date:	10/20/1976
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19340A146	List: ML19340A145 ML19340A149
References
BAW-1432, NUDOCS 8001100615
Download: ML19340A149 (14)

Text

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4 AMENDMENT.TO BAW-1432 OCONdE 3, CYCLE 2 'd?, II f ,

RELOAD REPORT

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1. Introduction This report amends the Oconee Unit 3, Cycle 2 Reload Report BAW-1432 dated June, 1976 to account for modifications to the core loading plan.

The revised core. loading plan for Oconee Unit 3, Cycle 2 will have four 2.64 wt % uranium-235 Mark B fuel assemblies replacing four 2.53 wt %

uranium 235 Mark B fuel assemblies originally intended to be inserted in Cycle 2.

The revised loading was selected because it yie'ds minimum perturbation to the core power distribution and shutdown margin. Core operational margins do not change. Loading four assemblies introduces no quadrant power tilt into the core. The inclusion of these four fuel assemblies does not affect the results of the core safety analyses or limiting conditions for operations as documented in BAW-1432. Therefore, the operation of Oconee Nuclear Station, Unit 3 at the rated core power of 2568 is fully justified.

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The following paragraphs describe the four replacement fuel assemblies and the effects of inserting these assemblies in Cycle 2.

l 2.. General Description Figure 1 is the revised core loading diagram for Oconee 3, Cycle 2. The initial enrichments of batches 2A, 2B, and 3 were 2.60, 2.67, and 3.00 we

% uranium-235, respectively. Batch 4A contains 52 fuel assemblies enriched to 2.53 we % uranium-235 and batch 4B contains four fuel assemblies enriched l to 2.64 wt % uranium-235. The nominal fuel assembly loadings are 463.6 kg t

of uranium for Batches 2A, 2B, 3 and 4A and 468.25 kg of uranium for' Batch 4B. Figure 2 is an eighth-core map showing the assembly burnup and enrich-ment distribution at the beginning of Cycle 2. Control rod group desig-nations are the same as shown in BAU-1432.

3. Fuel System Design-A comparison of the fuel rod design' parameters for Oconee 3 Batches 4A and 14B is shown in Table 1. The mechanical evaluation of the minor differences lLn fuel. rod design is discussed below.

Cladding Collapse The Batch 4B fuel rods are identical to the Batch 4A fuel rods in pellet density.and rod prepressure. These parameters indicate acceptable cladding collapse times as were determined for the Batch 4A fuel.

Cladding Stress The Batch 4B fuel rods are identical to the Batch 4A fuel rods in pellet density and rod prepressure. Therefore, the Batch 4B fuel rods are. completely equivalent with Batch 4A fuel rods from a cladding stress point of view.

l Fuel Pellet' Irradiation Swelling L The fuel design criteria specify a limit of 1.0% on cladding circumferential l

I plastic strain. The Batch 4B pellet design is set so thre the plastic cladding strain is less than 1% at 55,000 MWH/MTU. The conservatisms in this analysis are identical to those listed in Section 4.2.3 of BAW-1432.

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,- .s TABLE 1 FUEL ROD DESIGN PARAMETERS

-Component Oconee III, Batch 4A- Oconee III, Batch 4B Fuel Assembly Type Mark B4' Mark B4 235 2.64 Initial Fuel Enrichment, Wt % U 2.53

. Fuel Rods

~0D, Inch 0.430 0.430 ID,_ Inch 0.377 0.377 Fuel Pellets Diameter, Inch 0.3695 0.3697 Fuel Density, % TD 94.0 94.0 Undensified Active. Fuel Length, in. 142.2 143.5 a

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_ - - 1- _a Figure 1. Core f.oading Diagram tor Oconee 3, Cycle 2 F14EL TRANSFER w

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, 4 4A 4A 4B -

4A 4A

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, 4A 4A 4A 23 3 2a 4A 4A 4A

'F-7 A-8 F-9 4A 3 3 2s 2A .3 2A C 25 3 3 4A B-4 A-6 C-6 D-5 B-8 D-11 e-10 A-10 3-12 4A 3 28 3 3 4A 3 4A U 3 3 2B 3 4A D-2 E-8 A-7 R-5 C-4 3-11 A-9 H-11 D-14 4A 3 3 3 2A 28 2A 25, 2A 3 3 3 4A F-1 C-1 C-3 E-6 B-7 G-4. B-9

~~ E-10 C-13 C-15 F-15 4A 4A 28 3 2A 2A 3 2A 3 24 2A F-3 3 25 4A 4A E-2 F-5 D-7 B-6 C-8 B-10 C-12 F-11 E-14 F-13 4  ;

23 2A 4A

_ 4 2a 3 3 28 3 3 2n 4A 2A 28 4A C-6 E-4 G-2 F-2 D-3 G-8 l

C-12 F-14 C-14 E-12 ':-10 '

4B 3 3 3 2A 2A 2B 28 25 2A 2A 3 3 3 48 H

, H-1 H-2 N-3 N-7 N-3 H-7

, H-8 H-9 H-13 D-9 D-13 H-14

' H-15 4A 2B 2A 4A 2n 3 3 2s um K 3 3 2a 4A 2A 2n 4A K-6 M-4 K-2 L-2 0-4 K-8 N-13 L-14 K-14 M-12 K-10 i 4A 4A 2s 3 2A 2A 3 2A i

3 2A 2A L 3 28 4A 4A M-2 L-5 K-4 P-6 0-8 P-10 N-9 L-11 M-l'4 L-13 4A J 3 3' 2A 25 2A M 2B 2A 3 3 3 4A -i I

L-1 K-1 0-3. M-6 P-7 K-12 P-9 M-10 0-11 K-15 L-15 1 4A 3 28 3 3 4A j

" 3 .4 A 3 3 2s 3 4A N-2 H-5 R-7 .'P-3 0-12

-- P-11 R-9 M-8 M-14 4A 3 3 2B 2A 3 2A 1

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'P-4 R-6 0-6 N-5 P-8 N-11 0-10 R-10 P-12

.p 4A 4A 4A 2a 3 2u - 4A 4A 4A L-7 R-8 f. - 9

.R 4A 4A 4B 4A 4A

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- Figure 2. Enrichment and Barnup Distribution for Oconee 3, Cycle 2 8 9 10 11 12 13 14 15 l 2.67 2.67 2.60 2.60 3.00 3.00 3.00 2.64 13.205 17,759 17.079 17.528. 12.782 12.399 11.747 0 1

3.00 3.00 2.67 2.53 2.60 2.67 2.53 12.782 13.158 14.638 0 15.544 19.002 0

'l 2.60 2.60 3.00 2.67 2.53 2.53 17.528 17.714 12.199 15.653 0 0 i

y 3.00 3.00 3.00 2.53 9.174 11.254 8.652 0 2.67 3.00 2.53 N

19.018 8.577 0 2.53 C 0

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5. Nuclear Design The replacement of four 2.53 wt % uranium-235 fuel assemblies with four 2.64 wt % uranium-235 fuel assemblies will not have a significant effect on core nuclear performance. Analyses have shown that when the four replacement assemblies are loaded on the core majer axes as shown in Figure 1, the radial and total power peaks for the core over the revised reload cycle will not exceed those calculated with the original Cycle 2 loading. Beginning of cycle power distributions for the revised loading and the original submittal are shown in Figure 3.

Because of the small relative change in fuel enrichments and core power distribution, the core physics parameters are essentially unchanged from

! those documented in BAW-1432. Analyses have also shown that the total control rod worth and the maximum ejected and stuck rod worths are only slightly affected.

The shutdown margin calculation for the revised loading is shown in Table 2.

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~. ,m tr Table 2. Shutdown Margin Calculation for oconee 3, Cycle 2

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' BOC' , % Ak/k EOC(a) , % Ak/k

-Available Rod Worth

,, Total rod worth, HZPI) 9.77 9.70 Worth reduction due to burnup of -0. 9 -0.30 poison material Maximum stuck rod, HZP -2.27 -2.23 Net worth 7.31 7.17 Less 10% uncertainty -0.73 -0.72

, , Total available worth 6.58 6.45 Requirqd Rod' Worth

, Power deficit, HFP to HZP 1.65 2.30 Max allowable inserted rod worth 1.22 1.58 Flux redistribution 0.40 1.00 Total require'd worth 3.27 4.88 Shutdown Margin

-Total avail. worth - total. req. worth 3.31 1;57 Note: Required shutdown margin is 1.00% Ak/k.

(a)For shutdown margin calculations, this is defined as approximat ely

-- 226 EFPD, the latest time in core life at which the transient bank is .tearly full-in.

(b)HZP: hot zero power; HFP: hot full power.

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Figure 3. BOC (4 EFPD), Cycle 2 Two-Dimensional Relative Power Distribut ion - Full Power, Equilibrium Xenon, Normal Rod Positions (Croups 7 and 8 Inserted) j 8 9 10 11 12 13 14 15 H '1.30  !

1.19 1.15 1.11 1.18 0.83 0.78 0.63 1.29 1.18 1.14 1.11 1.18 0.83 0.79 0.66 7

K 1.19 1.33 1.30 1.18 1.22 1.18 0.55 0.70 0.62 1.33 1.29 1.18 1.27 0.55 0.71 0.63 8

L 1.15 1.30 1.12 1.09 1.12 0.96 1.00 0.55

1.14 1.29 1.12 1.09 1.12 0.96 1.00 0.56 l

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[ 1.11 1.18 1.09 1.34 1.31 1.28 0.97 l.11 1.18 1.09 1.34 1.30 1.28 0.97

, 8 1.18 1.22 1.12 1.31 1.18 1.12 1.14 0.70 1.22 1.12 1.30 1.12 1.14 0.69 i

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O.83 0.55 0.96 1.28 1.14 0.77

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0.83 0.55 0.96 1.78 1.14 0.76 i

! P 0.78 0.70 1.00 0.97 0.70 0.79 0.71 1.00 0.97 0.69 R

0.63 0.62 0.55 4

0.66 0.63 0.56 w

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X.XX - Relative Power Density with Original Cycle 2 Loading i Y.YY Relative Power Density with Revised Cycle 2 Topling C t

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i i-6; Thermal-Hydraulic' Design The B'atch 4B-fuel. assemblies are equal to or better than the Batch 4A fuel

• l assemblies from a thermal-hydraulic view point. Hydraulically, the as-semblies are identical so there will be no effect on core flow distri-

[ .bution. Batch 4B has a slightly larger pellet diameter, the same theoretical i .

l- density, and a slightly. longer active fuel length. This will result in the Batch 4B fuel assemblies having a higher fuel melt limit, higher DNB margin and a lower densification penalty. For these reasons the fuel- assemblies l in Batch 4B will not limit the performance of the core.

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7. Accident and Transient Analysis The changes in the Nuclear and Thermal-Hydraulic Design discussed in Sections-5 and 6, respectively, have been evaluated for effect on the transient analysis. However, due to the fact that the core physics parar.eters are essentially unchanged and that the fuel assemblies in Batch 4B are not thermal-hydraulically limiting, the results and conclusions ,

in Section 7 of BAW-1432 remain valid.

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8. Proposed Modifications to Technical Specifications The revised loading plan will not affect the technical specifications from I BOC to 115 EFPD and will not affect the limiting conditions for operation at power throughout the cycle.- However, after 115 EFPD, the rod position

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11mics' for shutdown margin must be modified to account for the minor changes in rod worth. The limiting condition for operation at power is determined by the LOCA kw/ft rod position limits and not the shutdown margin rod' position limits. Changes to the shutdown margin rod position limits l are presented in Figures 4, 5, 6, and 7.

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Figure 4.

Oconee 3, Cycle 2 - Rod Position Limits for Four-Pump Operation From 115 (t10) EFPD to 226 (!10)

EFPD 122 ,102 170, 102g 100- Operation in this g'209.4, 102 Region is Not 90 . Allowed 170, 9 -

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Oconee 3, Cycle 2 - Rod Position Limits for Four-Pump Operation After 226 (110) EFPD Operation in this 149 102 255 102 100' -

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Oconee 3, Cycle 2 - Rod Position Limits for Two- and Three-Pump Operation From 115 (t10) to 226 (110) EFPD Operation in this 122 102 46,102 218 02 Restricted

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Oconee 3. Cycle 2 - Rod Position Limits for Two- and l Three-Pump Operation After 226 (t10) EFPD i

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