ML19322C154
| ML19322C154 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/31/1976 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| BAW-1424, NUDOCS 8001090564 | |
| Download: ML19322C154 (17) | |
Text
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January 1976 pyn**,? ?.T U pg g .3 iu r.wa-{.}'f I n0011 BB I [$$ EI}kb . i I _ i 4 I IPJADIATION OF TkV 17x17 DDiCNSTRATION I I 15SDGLIn., IN OCONIE 2, CICLE 2 !
- Re1.oad Report - ,
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I . I - I ; 2 I ! ! I ' I : l ) EETUM TO EULATG3Y CE5?J1 F.23 i E003 018.- . . . . . . . ...,
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;, ' IRRADIATION 07 TWO 17x17 DD80hTRATION y ASSD(BLIES IN OCONEI 2. CYCLE 2
- Reload Report -
L[ l I l , I l by I P. C. Childress {;i
. J. J. Woods T. U. Ake
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- BABCOCK & WILC0K Power Generatien Group
., Nuclear Power Generation Division j ', P. O. Box 1260 Lyr.chburg, Virginia 24505 4
- i Babcock &Wilcox
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'! CONTENTS L' t Page l
l 1. INTRODUCTION . . . . . . . . . ... . . . . .. .. ....... I 3
- 2. MECHANICAL DESIGN . . .. . .. . . . . . . . ..O....... 2 l I 7
- 3. NUCLEAR DESIGN . . . . . ... .... . ... . .........
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- 4. THEFMAL-HYDRAULIC DESIGN . . . ... .. .. ...
J, , EVALUATION OF D.,fFERENCES . . . ... . .. . .. ........ 12 5. g.
'( DISTRIBUTION . . . .. ... . .... ... - * ******* 13 d List of Fleures Figure 1 ,r Fuel Assembly Comparison . .... .... ... ....... 5 1.
End Fitting Desigr.. . . . . .. .. . .. ..... ...... 6
- 2. 8
- 3. Fuel leading Pattern for Oconee 2, CycD 2 . .........
, 4. Relative Power Densities for Oconee 2 - Beginning of Cycle 9
- 2. Transient Rods In .. . .. .. . ... . .........
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N lL L i i !~ g 1. IXTRODUCTION su
- i j Duke Power Company will irradiate two 17x17 fuel rod essemblies in the Oconee Unit 2 reactor during fuel cycles 2, 3, and 4. This irradiation will be a demonstration of the in-reactor performance of the 17x17 Mark-C fuel assembly.
This report identifies the differences to be expected between a core opet sting ! with IEl 15x15 fuel assemblies and one operating wilh 175 15=15 fuel assemblies and two def;Dnstration 17x17 assemblies. An evaluation of thesa differences indicates that reactor safety and performanc~ are not adversely af fected by l the presence of the two demonstration assemblies. Ll ; u i
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- 2. MECHANICAL DES E4 L.
i ( ,~ The two 17=17 demonstration assemblies to be irradiated in Oconee C are struc-1 turally identical, to the extent possible. to the generic Mark-C design. The
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l end fittings and spacer grid elevations were modified to be compatible with the 15x15 fuel assembly (Mark-B) interface constraCes. l Tb Hark-C demonstration assembly is a 17x17 array of fuel reds wii:h th- cc-external envelope as the Mark-B. Table 1 is a dimensional comparison of I ll Mark-B and Mark-C demonstration assemblies. The Mark-C dcmon;tration assen-lfb 1s blies are mechanically compatible and interchangeable with Mark-B Miles
, with the exception of the control comp w nt interface.
i The demonstration assemblies have Mark-B upper end fitting plenums to ensure
< j compatibility with core internals and Oconee handling equipment (see Figures l'd 1 and 2). The design of the demonstration assembly apper end fitting necessi-tated the use of a single helical holddown spring instead of the four used in j the standard Mark-C design. The demonstration assembly holddown spring i.
rized to generate a compressive force greater than that of the single Mark-B l} spring and similar to that of the four Mark-C springs. l-L 'I The spacer grid elevations in the demonstration assemblies are identical to l I 'i
- those of the Mark-B assemblies. This is required to provide lateral interface support from grid to grid or grid to baffle. The standard Mark-C fuel assen-t , blies will have slightly different spacer grid elevations than the demonstra-l tion assemblies to make assembly insertion and withdrawal easier.
[J i Standard Mark-C fuel rods are used in the two demonstration assemblies. ~The
! pellets and cladding have been examined extensively and their properties .I !
cataloged for subsequent reference. Two lengths of fuel pellets have been j manufactured for the demonstration assemblies. All other pellet parameters i are identical. Utilitation of different fuel pellet length-to-diameter ratios j a allows different fabrication and loading techniques to be investigated. The larger pellet L/D ratio is similar to the L/D successfully used in the Mark-B t i j '- { BabcC;:k s.WHcox
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' fuel assently. The smaller L/D is expected to improve the performance of the 17 x 17 demonstration fuel assembly, o
The demonstration assemblies are compatible with the Oconee incore detectors. j There is a minimum diametral clearance of 0.096 inch between the instrument
tube and instrument probe. This clearance is more than adequate to allow free movemect of the instrument probe in the instrim-ne tube.
U The Mark-B control rod assemblies at Ocence are not compatible with the Mark-C 1i demonstration fuel assemblies. This dictates that the demonstration assemblies
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" be lo sted in non-control rod positions. Design limitations also preclude the use of a Mark-C orifice plug assembly in the demonstration fuel assembly. There-
, I' j U fore, the nuts that attach the demonstration assembly guide tubes to the upper i,, end fitting have been orificed to provide the same 1ypass flow as a standard I r*
Mark-C guide tube with orifice plugs inserted. ld i The two assemblies have been extensively precharacterized and will be examined I d af ter every cycle. Observation vindows have been cut into the end spacer grids to allow measurement of fuel rod growth without removing the rods from the fuel asserLly. The st.atie and dynamic structural characteristics of the demonstration assee-l, blies are compatible with the Mark-B assemblies. The demonstration assemblies have been designed to maintain their mecharical integrity through three cycles of opesation and successfully withstand all seismic and LOCA loads postulated
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e Table 1. Oconee 2, Cycle 2 Fuel Assembly Comparison Mark-C l Mark-B demo
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I Total No. of assemblies 175 2 > w s E. of fuel rods / assembly 208 264 No. of guide tubes 16 24 If (j ; No. of instrument tubes 1 1 !! "* Fuel rod OD, in. 0.430 0.379 Cladding-per.et diam. gap, in. 0.007 0.008 Fuel rod cladding thickness, in. 0.0265 0.0235 Fuel pellet diameter, in. 0.370 0.324 Fuel pellet length, in. 0.700 0.600(b) , 0.375(b) Ii.
- t fuel pellet density, I TD 92.5, 93.5I ") 94.0 i}
(, l! t I Fuel rod pitch, in. 0.568 0.502 '. ! u . Fuel assembly pitch, in. 8.587 8.587 f' 144.0, 142.5I *) t 1 Nominal active fuel length, in. 143.0 fl ,, ll Hot flow area, in.2 40.39 40.35 l t ' Avg linear heat rate, kW/ft 5.81 4.61 [- 1 j Heat transfer area / bundle (hot), f t2 281.5 312.7
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- Batch 4, Oconee 2.
l' i t O)0ne assembly with 0.375-inch pellets only. One assembly has 11 fuel ,f
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j; rods with 0.375-inch pellets while the remaining rods have 0.600-i ~ inch pellets. } f <,
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Ficure 1. Fuel Assenbly Co ,pariscu y- 91 ,,
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ASSEMBLY FUEL ASSEMBLY _ n=henck n.wiscar
C __\ I I Figure 2. End Fitting Designs aw o u ers
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L i u iU Ii 3. NUCLEAR DESIGN iL
< t j ] The Oconee 2. Cycle 2 loading pattern is shown in Figure 3. The 17=17 demon-stration assemblies will be loaded in synanatric core locations A-6 and R-10.
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! They will have the same enrichment, 2.64 wt 1 2350, as the other batch 4 fuel assemblies. Figure 4 shows the core power distribution at the beginning of Cycle 2 calculated for a core loaded entirely with 15=15 assemblie.=.
JI The nuclear characteristics of the 17=17 and 15=15 assemblies are nearly iden- [ tical. Initially, the average relative power densities in the 17=17 assembifes. l will be slightly less than those in the symmetric 15=15 assemblies. The reac-l~ l l tivity of a 17 17 assembly is approximately 0.2% less than that of a 15=15
- assembly of the same enrichment: this is a result of differences in fuel load-g ing. The 17=17 and 15=15 assemblies contain 456.1 and 463.6 kg of uranium, respectively. The slightly lower relative power density in tl e 1717 assemblies results in a slightly lower burnup, which tends to reduce the reactivity dif-l a j ferential and increare the relative power density. Trie relative power density in the 17=17 assemblies approaches that of the synsmetric 15=15 assemblies as the cycle progresses.
( , 4 The peak pin-to-average assembly power in a 17x17 fuel assembly is approxi-
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l'f a mately 1.0 to 1.3% lower than that in a 15=1.5 assembly of the same enrichment in i similar core location. The reduction is primarily due to the larger k [ number of water holes (control rod guide tubes and instrument channel), which
'I i" are arranged ~)re uniformly over the assembly.
l.. There are no significant differences betwecn the Doppler and moderator tempera-
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{~ ture coef ficients for 17=17 and 15=15 assemblies of the same enrichment. There-fore, the presence of the 17=17 fuel assemblies will not discernably affect l'? overall core reactivity coefficients or performance.
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u i i ; i s. v Figure ?. Fuel loading Pattern for Oconee 2. Cycle 2 l ks . s 8
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t. I 4. THERMAL-HYDRAULIC DESIGN L-
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L. Analytical results show good thermal-hydraulic compatibility between the demon-stration assemblies and the remainder of the core. The similarity in flow areas and hydraulic resistance of the 15=15 and 17=17 assemblies will allow both to operate safely in the same environment. The pressure drop loss coef ficients for the demonstration assemblies were exper-inentally determined in a heated-water test facility that simulated reactor i j conditions. The test results showed that the overall hydraulic resistance of the test assemblies is greater than that of a Mark-B assembly. Because of I; this dif ference in resistance, the Mark-B assemblies will receive slightly more flow than they would in an all-Mark-B core, whereas the Mark-C test assemblies will recieve slightly less flow than would Mark-B assemblies in the same core L2 locat ions. The effect of this extra resistance on the total core pressure drop is negligible. A lif t analysis was performed to demonstrate that the Mark-C test assemblies
. will not lift during any expected reactor conditions. The results for the I
) Mark-C demonstration fuel assembly showed a substantial margin in holddown force even at reactor system flow rates 20% greater than expected. This anal-
- ysis used several conservative assumptions concern *ng fuel assembly resistance.
l flow, and holddown spring tension. The 17x17 assemblies will have a negligible t I t effect on the lift margin of the Mark-B assemblies, u i The thermal performance of the Mark-B fuel assemblies in the presence of the
* ' two demonstration assemblies will be equivalent to or better than their per-
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formance in a core composed entirely of Mark-B fu al assemblies because of the { slightly increased Mark-B fuel assembly flow. Thus, the Mark-B fuel assembly I minimum DNER will be somewhat improved over the reference value reported in j the FSAR. In addition, a DNBR analysis was performed for the Mark-C demon-d stration assembly using the BAW-2 correlation with conservatisms and i E _m_ sabcock awiscom
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l\ 'L iI L uncertainties associated with the reference " maximum design case" as noted in the FSAR.I These results showed that the ainimum DNBR, 3.23, ws well above [ the 1.55 design minimum. g Analyses have been performed to determine the maximum linear heat rate that will L cause cen'6er11ne fuel melt in 17x17 and 15x15 assemblies. nese results show that cents.rliae melt would occur at apprcximately the same linear heat rate for ,I, y both assembly types, assuming densification and using NRC-imposed restrictions on the TAFY-3 computer code.2 The average linear heat rate for the 17x17 fuel assembly is reduced approximately 25% due to the extra fuel length; therefore, u 17x17 demonstration fuel assemblies will operate at a lower temperature than the 15x15 assemblies, and the demonstration fuel assemblies will not be limit- [ ing with regard to the centerline fuel melting criterion.
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4 J I Final Safety Analysis Report Duke Power Company, Oconee Nuclear Station, Docket No. 50-270.
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2TAFY - Fuel Pin Temperature and Cas Pressure Analysis, BAU-10044, Babcock & Wilcox, April 1972.
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k L c EVALUATION OF DIFFERENCES (. 5. U t I u The two 17x17 fuel assemblies have a substantial additional operating margin j beyond that of the 15x15 fuel assemblies with respect to fuel temperature lim-l
" itations. Together with the lower average linear heat rate, the placement of the demonstration assemblies in the core further guarantees that the 17x17 l
d assembly will not restrict the operation of the core. As described in section 3, the core power distribution remains very nearly un-
; - sffected by the presence of the two demonstration assemblies. Minor local a
1 ,, reactivity perturbations do occur, but their ef fect is negligible tecause there are only two 17x17 assemblies out of a total of 177 fuel assemblies in the core. The total fission product inventory of each 17x17 assembly is expected to be nearly identical to that of a 15x15 assembly in the same core location; how- ( 1
,f ever, because the 17x17 assembly operates at a significantly lower average linear heat rate, the fraction of the activity available to the fuel pin gap f
and plenum is conservatively lower than that expected with the 15x15 fuel
;j design.
Therefore, the loading of two 17x17 Mark-C demonstration assemblies in Oconee
; 2, cycle 2 will not discernably affect the nuclear, suchanical, or thermal-hydraulic character of the reactor, nor will it adversely affect the existing lll safety analysis.
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