ML19329D678
| ML19329D678 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/20/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329D677 | List: |
| References | |
| NUDOCS 8003160299 | |
| Download: ML19329D678 (6) | |
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N Evaluation of Crystal River, Unit 3 Replacement Fuel Assembly 3A33
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' Background Damage incurred to Crystal River, Unit 3 fuel assembly 3A33 during handling necessitated replacement with a new assembly.
The replacement 3A33 assembly is made up of 40 fuel rods removed from the original assembly and 168 replacement fuel rods.
On March 26, 1976, Babcock and Wilcox submitted a report (Reference 1), which described the replacement fuel assembly along with analyses performed by B&W on the potential for cladding creep collapse, fuel / clad interaction, fuel densification and fuel swelling.
In response to requests by the NRC staff for additional information, Unis report was supplemented by a subsequent submittal on May 4, 1976 (Reference 2).
Description of Replacement Fuel Rods The only significant differences between the 168 replacement fuel rods in replacement assembly 3A33 and the r'emaining fuel rods in the first cycle leading of Crystal River, Unit 3 relate to the internal spacers and the
. fuel pellets.
Spring spacers and Zircaloy tubular spacers replaced the corrugated tube spacers and zirconia ceramic spacers in the original rods.
The newer types of spacers are representative of those currently in operation in several B&W reactors and thus are not of concern. -The fuel j
. pellets, however, differ slightly in enrichment, density, and active length, as shown in Table I.
Of.particular interest to the Regulatory staff were the pctential effects of the-low-density, 90.c% theoretical density (TD), ~ pellets on ther=al and mechanical performance.
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TABLE T
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COMPARISON OF FU L I'ARAMETERS I
NO.
FUEL-ASSEM2LY-FUEL RODS ENRICHMENT
%TD STACK LENGTH, IN.
W/% U 235 1
e-3A33 40 1.93 92.5 144 s
s 1
168.
1.98 95.35 23-1/2 (Upper Zone) 1.94 90.9 95-3/4 (Central Zone) > 142-3/4 1.98 95.35 23-1/2 (Lower Zone) s Remaining.
Q Assemblies 208
-1.93 92.5 144 in Batch 1 4
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t Summary of Re;ulatory Evaluation Analyses of the thermal-hydraulic performance of the replacement
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fuel rods in replacement' assembly 3A33 were performed by B&W and reported in the March 26, 1976, letter (Reference 1).
A co=parison of the results of these analyses to analyses of the 40 original fuel rods in assembly 3A33 and the fuel rods in the remaining 176 fuel assemblies is shown in Table II.
The results of the analyses show that, except for the engineer-ing hot channel factor, all thermal-hydraulic performance parameters for the replacement fuel rods in fuel assembly 3A33 are not more restrictive than for the fuel rods in the limiting fuel assembly in the remaining 176 fuel assemblies.
The effects of the higher engineering hot channel factor for fuel assembly 3A33 were not addressed under " Fuel, Mechanical Design," but were_ treated under " Nuclear."
Cladding creep collapse analyses vere performed by B&W in accordance vi.th material properties and design procedures set forth in Topical
_ Report B&W-10084P-A, entitled " Program to Determine In-Reactor Performance of B&W Fuels" (Reference 3).
The evaluation was completed using the NRC -
approved CROV creep ovalization analysis code described in Section 3 of the cited repore.
In addition other conservations were introduced, as described in reference 2.
Results of the analyses indicated a collapse time > 14,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, compared to the required 10,320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br /> associated with the single cycle burn of assembly 3A33.
Pellet / cladding mechanical interaction (PCMI) and fuel swelling effects were addressed by.B&W in their cladding strain analysis.
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Of the pellet densities used in assembly 3A33, the 90.9% T.D. pellets
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represent the limiting PCMI case at the peak pellet burnup seen by the assembly. Accordingly, cladding strain analyses were performed on the i
90.9% T.D. fuel corresponding to the worst-case specification dimensions and.the as-built, 20 dimensions.
The analyses were performed in accordancu with material data and design models set forth in Section 3 of Tcpical Report B&W-10054, Revision 2, entitled " Fuel Densification Report" (Reference 4). This represented the same approach as used in the Crystal River SAR except 6nat additional conservations were introduced for the 3A33 analyses as listed in reference 2, p. 2.
The results of tha analyses i
indicated that the total circumferential strain resulting from PCMI for j
the worst-case-specification analysis and the as-built dimensions analysis were 0.80% and 0.48%, respectively, as compared to the B&W " design" value of 1.42%.
In su= mary, the mechaulcal design and thermal analysis aspects of the 168 replacement fuel rods in Crystal River, Unit 3 replacement fuel
.m assembly 3A33 have been analyzed by ESW, using NRC-approved codes and methods
-and in accordance_wiri material data and design models approved by NRC.
Evaluations of-the potential for cladding creep collapse, pellet / cladding j
mechanical interaction, fuel densification and fuel swelling were made.
The result's of these analyses have shown that the 3A33 fuel rods are within acceptable design limits for first cycle operation of Crystal River, Unit 3.
The information provided on the results of the analyses of the 3A33 replace-ment fuel assembly provides an acceptable basis for demonstrating their adequacy.
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.: e References 1.
J. T. Rodgers, Asst. V.P., B&W, to D. A. Butler, Chief, LWR
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Research #4, " Report on Replace =ent Assem*oly 3A33," March 26, 1976.
2.
J. T. Rodgers to D. A. Butler, " Supplement to Report on Replacement Assembly 3A33,."
May 4, D76.
3.
A. F. J. Eckert, et al, " Program to Determine In-Reactor Performance i
of B&W Fuel.4," B&W-10084P-A, Nov. 1974.
4.
R. A. Turner, " Fuel Densification Report," B&W-10054, Revision 2, May, 1973.
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CL,PARISON OF Ti!ERMAL-HYDRAULIC
.., PARAMETERS __ _.
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168 FEFLACEMENT 40 ORIGINAL FUEL RODS FUEL RODS IN IN'3A33 AND FUEL RODS FUEL ASSEMBLY IN REMAINING 176 FUEL THERMAL-ilYDRAULIC CRITERA 3A33 ASSEMBLIES 1.
Linear lleat ute Limit Based on Centra'. Fur ~ Melting, KW/Ft.
For Fuel Density:
- a. 95.35% TD 21.46 s.
- b. '90.9% TD 19.96
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- c. 92.5% TD 19'.7 2.
Average Linear !! cat Rate, KW/Ft.
5.765
,[5.771 3.
Average Fuel Temperatures (Stored Energy), -
- a. At Average Linear lleat Rate:
(1) 95.35% TD 12,85, (2)~ 90.i,3 TD 1327
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J (3) 92.5% TD 1335 b.
At 18'KW/Ft.
(1) 95.35 ' TD 28/0 (2) 90.9% TD 3066 1
.(3) 92.5%
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Engi.'.ccring !!ot Channel Factor 1.026 1.014 5-DNbR Fenalty Due to Fuel Danstfication,%* 1.9 2.9 9
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