ML20077K065
| ML20077K065 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/30/1991 |
| From: | Miller R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19302E970 | List: |
| References | |
| WCAP-12997, NUDOCS 9108060290 | |
| Download: ML20077K065 (21) | |
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wr.snmmuusn nonarrov ciuo WCAP-12997 I I I Au0rnentation Factor Elimination for Westinghouso Fuelin Fort Calhoun I I P. J. 1;erating June 1991 I APPROVED: I R. S. Miller, Manager Fuel Performance Technology
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I I I I Westinghouse Electric Corporation Commercial fluclear Puel Division P. O. Box 355 I Pittsburgh PA 15230-0355 0 Westinghouse Electric Corporation 1991, All Rights Reserved I
I Table of Contents Table of Contents . . . . . . . . . . . . . . . . . . . 11 List of Tables . . . . . . . . . . . . . . . . . . . . iii List of Figures . . . . . . . . . . . . . . . . . . . . iii 1.0 Summary . . . . . . . . . . . . . . . . . . . . . 1 I 2.0 Introduction . . . . . . . . . . . . . . . . . . . 1 3.0 Review of Axial Gap Data for Wentinghouse Fuel . . 2 1 3.1 Gamma-Scan Data . . . . . . . . . . . . . . . 3.1.1 General . . . . . . . . . . . . . . 3 3 3.1.2 Zion Program . . . . . . . . . . . 4 3.1.3 Surry 17X17 Program . . . . . . . . 5 1 3.1.4 BR-3 liigh Burnup Program . . . . . 5 3.1.5 Zorita Program . . . . . . . . . . 5 I 3.2 3.3 3.1.6 Gamma-Scan Data Evaluation . . . Flux Map Data . . . . . . . . . . . . . . . . Other Data . . . . . . . . . . . . . . . . .
. 6 7
8 4.0 Conclusions . . . . . . . . . . . . . . . . . . . 8 5.0 References . . . . . . . . . . . . . . . . . . . . 9 I I I I I I 11
s List of Tables L Table 3.1-1 On-Site Gap Measurement Data from the f Zion Extended Burnup Program . . . . . . 10 l Table 3.1-2 Ilot Cell Gap Measurement Data from the Zion Extended Burnup Program . . . . . . 11 J Table 3.1-3 On-Site Gap Measurement Data from the Surry 17X17 Extended Burnup Program . . 12 Table 3.1-4 llot Cell Gap licasurement Data from the I Table 3.1-5 Surry 17X17 Exter.ded Burnup Program . . llot Cell Gap Measurement Data from the 13 BR-3 Extended Burnup Program . . . . . . 14 Table 3.1-6 Gap Measurement Data from the Zorita 1 liigh Burnup Program . . . . . . . . . . Incore Fuel Stack lleight Data From 19 15 Tabic 3.2-1 Pre-Pressurized Fuel Regions . . . . . . 16 List of Figures i Figure 3.1-1 Figure 3.1-2 Uncorrected Axial Gaps as a Function of Axial Elevation . . . . . . . . . . . . Corrected Axial Gaps as a Function of 17 Axial Elevation . . . . . . . . . . . . 18 l l I I iii
I 1.0 Summary current go'. oration Westinghouse fuel is highly stable with I respect to fuel densification. measurer.cnt of axial gaps which form in the fuol column. The primary source of this data is 1-scanning of irradiated This has been verified by Incore flux traces have also been examined to I fuel rods. detect flux spikes which could occur if larger a>:ial gaps were present in the fuel column. These data confirm that axial pellet column gaps due to Iuol densification are very infrequent and extremely small (essentially undetectable with Incore flux trace systems). I In fact, the small gaps which are observed by y-scans are completely consistent with the gaps which could occur due to differential contraction between the fuel and the cladding associated with the change irom end of cycle full power I conditions to the cold shutdown moacurement condition. corrected for differential contraction, the maximum observed When axial pellet gap is less than [ J' inches. Gaps of this *(a,b,c) magnitude do not significantly impact local power peaking. It is concluded that an augmentation factor of 1.0 in appropriate for the Westinghouse fuel to be supplied for the I Fort Calhoun reload cycle designs, since significant axial gaps do not occur in current generation, stable Westinghouse fuel. I Introduction 2.0 The augmentation factor, or spike factor, was developed in response to early observations of significant fuel densification, which resulted in the formation of axial gaps I in the pellet column during irradiation. In some of the early commercial nuclear plant first cores, fuel was intentionally fabricated with low initial densities by intentional undersintering during manufacture to generate substantial porosity which was believed to be necessary to offset expected fuel swelling. The resulting pellet I microstructure produced excessive densification during operation, with both radial and axial pellet shrinkage. At the same time, a number of these early fuel regions were not pre-pressurized, so that there was substantial early I creepdown and contact with the fuel. The axial shrinkage associated with densification, coupled sith the early fuel-clad contact in unpressurized fuel rods, resulted in the formation of significant axial gaps in the fuel column. One consequence of these large axie.1 gaps was local flux peaking, as detected by the Incore flux monitoring system. I power peaking occurs because of the reduced neutron f I
absorption in the gap relative to that observed when fuel is present. Clad flattening into the gap further increases the local power peaking because of the increased local moderation. The augmentation f actor, or spike f actor', was developed to bound the effects of densification rn power peaking. The current augmentation factor was conservatively derived assuming maximum gap size and worst case gap I distribution. In response to the early observations of significant I densification, the fuel fabrication process was modified to produce higher initial densities with increased sintering temperatures and times. These early process changes I resulted in stable fuel with respect to denuification. Substantial development effort has been opent cince the time these process improvements were made to better understand the fuel microstructure characteristics which control the I densification mechanism. These studies identified small grain size and excessive small porosity as the most significant factors affecting densification behavior. Current generation fuel fabrication processes have been developed to consistently produce fuel pellets which exhibit controlled microstructure with respect to both grain size I and pore size distribution. As a result, current Westinghouse fuel continues to be stable with respect to densification, and significant axial pellet column gaps do not occur. I Review of Axial Gap Data for Westinghouse Fuel 3.0 Westinghouse has accumulated a substantial fuel performance data base which supports the climination of the augmentation factor for the Westinghouse fuel to be supplied to Fort I Calhoun. The most significant data base for evaluation of axial fuel column gaps is the availability of y-scan data for over 360 fuel rods (including both early densifying l fuel, carly stable fuel, and current stable fuel). The y-scan da*.a base representative of current Westinghouse fuel is presented in Section 3.1. The Incore flux trace measurement system can also be used to detect flux spiking I associated with significant axial gaps. Section 3.2 presents a review of an early evaluation of Incore flux trace measurements which were collected for 19 pressurized I fuel regions. Both early densifying fuel regions and regions fabricated essentially like current fuel fabrication practice are included in this evaluation. Other sources of I data on fuel densification behavior include direct hot cell measurements of irradiated fuel density and visual examinations from several on-site fuel examination programs. I I I
3.1 Gamma-Scan Data e 3.1.1 General L Gamma-scanning of irradiated fue1 rods has been used to determine gap frequency, size and axial location as well as to measure the total fuel stack length. These data have been collected from both on-site fuel examination campaigns and from hot cell programa. A substantial data base on fuel stack lengths and axial gaps was accumulated as part of early investigations of fuel densification, as reported in Reference 1. Axial stack I length data, corrected for axial gaps, were presented for 284 fuel rods. This data base included fuel rods irradiated in the Saxton Core III, the Zorita reactor in Spain, the Deznau Unit I reactor in Switzerland, and the Point Beach I Unit 1 reactor. Most of these rods, however, contained fuel fabricated at low initial density and with relatively low sintering temperature, resulting in a highly densifying l fuel. These data are not representative of current fuel, but they provided conservative information on gap distributions which was then used in the calculation of the I spike factor. 3 Additional y-scan data has been collected for 58 fuel rods that are representative of current generation pre-l preocurized rods fabricated with non-densifying fuel. Six of these rods were scanned after more than one cycle, and two rods wero scanned both on-site and at the hot cell, so I that there are a total of 66 7-scan traces which provide data representative of current Westinghouse fuel. These data have been obtained from extended burnup demonstration I programs conducted at Zionu As , Surry", the BR-3 reactor at Mol, Be lg ium', and the Jose Cabrera (Zorita) reactor in Spain'. Data obtained from each of these programs is summarized below. An evaluation of the y-scan data base I follows the summaries of the data obtained from each of these programs. Gamma-scanning provides an accurate means for measuring the axial gaps in the fuel column at the conditions of measurement. However, these measurements are performed on-site in the spent fuel pool or in the hot cell at I temperatures substantially less than experienced during actual operation. Axial gap formation can occur as a result of expected differential contraction as the rods go from the I end of cycle hot condition to the lower temperature measurement condition. l 3 I
I The measured axial gap data can be corrected to account for this expected differential contraction effect. This gap correction is calculated based on the predicted, axial dependent fuel and clad average temperatures at the hot -I condition at end of cycle using the NRC approved Westinghouse fuel performance models*. Actual rod average I power histories and axial power distributions are used to predict the fuel and clad temperatures as a function of axial position. The differential thermal expansion is
)
calculated for each axial interval, and is accumulated as a I function of height. The cumulative ditferential thermal expansion is then subtracted from the total cumulative gap , at each axial clovation to obtain the corrected hot gap. I ~ Since the uncorrected measured gaps in the BR-3 and Zorita rods were very small, no corrections were made to these rods. The BR-3 rods operated at very high power IcVels, and it is reasonable to assume that the observed gaps are 'I consistent with differential thermal expansion effects. Similarly, the maximum single gap observed in the Zorita I rods was only 0.008 inches, which is small relative to expected differential thermal expansion. 3.1.2 Zion Program In the Zion extended burnup irradiation program, four Westinghouse 15X15 fuel assemblics were irradiated for five cycles in Zion Unit 1 and Unit 2. Two of these assemblies were of the removable rod design, so that interior fuel rods I could be removed and examined on-site. Gamma-scanning was performed on-site for selected rods after one and two cycles of operation. In addition, some of these rods were permanently discharged at the end of each cycle to allow for I more extensive examinations in a later hot cell program. A two phase hot cell examination program was ccaducted on I these Zion fuel rods at the Battelle Memorial iwtitute. In the first phase, individual rods which had been discharged after one, two, three, and four cycles of operation were I examined in the hot cell along with a three cycle fuel assembly skeleton. In the second phase, a complete five cycle Zion fuel assembly was sent to the BMI hot cell and disassembled, allowing detailed examination of several individual fuel rods. The hot cell examination scope included y-scanr.ing of several of these rods. I Tabic 3.1-1 summarizes the gap measurements for the Zion fuel rods examined on-site. All of the observed gaps were small, with the maximum individual gap size of 0.103 inches
,3 occurring in one-cycle rod 687. Table 3.1-2 summarizes the j interpellet gap measurements obtained for the Zion fuel rods 4
I
I examined at the hot cell. The maximum individual gap observed for all of the rods examined in the hot cell was *(b,c) ( )* inches, which occurred in five cycle rod B15. 3.1.3 Burry 17X17 Program In the Surry 17X17 demonstration program, two Westinghouse 17X17 fuel assemblies were irradiated in the surry Units 1 and 2. These assemblies were of the removable rod design, and some rods were removed for examination on-site, including y-scanning, after one and two cycles of operation in a joint program with EPRI. In a subsequent DOE sponsored program involving the same demonstration assemblies, one of I these assemblies was irradiated for four cycles to an assembly average burnup of 42,300 MWD /MTU. In this program, twelve fuel rods (eight four cycle fuel rods and four three I cycle fuel rods) were shipped to the BMI hot cells for detailed examinations, including 7-scanning. I The interpellet gap measurement data obtained from the Surry 17X17 demonstration rods in the on-site measurements are summarized in Table 3.1-3, and the hot cell gap measurement results are shown in Table 3.1-4. I Interpellet gaps were observed in only two of the Surry rods measured on-site, and the maximum observed individual gap was only 0.097 inches. No interpellet gaps were observed in any of the rods I examined in the hot cell. 3.1.4 DR-3 High Durnup Program In the BR-3 irradiation program, five highly characterized 15X15 fuel rods were irradiated at lead rod power conditions I to rod average burnups near 60,000 MWD /MTU, with peak pellet burnup exceeding 73,000 MWD /MTU. These rods were shipped to the CEN hot cell under a joint DOE / Westinghouse program for I detailed examinations, including y-scanning. Table 3.1-5 summarizes the gap measurement results for these rods. No gaps were observed in the central higher power portion of I these rods, and only very small gaps were noted at the extremes of the rods. The maximum cumulative gap for these rods was only 0.012 inches, and this was the sum of numerous very crall interpellet gaps. No attempt was made to I evaluate the individual gap data for these rods. 3.1.5 Zorita Program The Zorita Research and Development Program was initiated in 1966 to demonstrate improve fuel rod design capability for I high-power, high burnup applications in PWR reactors. 'I hi s I i I
program was performed under a tripartite agreement between F Westinghouse, the Junta de Energia Nuclear (J.E.N.) and Union Electrica Madrilena, S.A. In this program, 164 removable fuel rods (96 non-pressurized and 68 pressurized) were inserted into f our removable rod ascelablics. The fuel in those rods was fabricated with a range of sintering temperature (1492*C to 1750*C) and initial density (93.1% to 94.8% T.D.) to permit evaluation of the impact of these
~
parameters on fuel performance. Detailed on-site and hot cell examinations were performed on several of these removable rods, including 7-scanning. Table 3.1-6 summarizon the gap moacurements for three Zorita rods which were fabricated within the range of the current Westinghouse i fabrication specifications for density and sintering temperature. No gaps were observed in two of these rods and one very small gap (0.008 inches) was noted in the third rod. I 3.1.6 Gamma-Scan Data Evaluation A total of 66 y-scan traces have been evaluated, from 58 fuel rods which are representative of current generation Westinghouse fuel. Six of the Surry rods which were I examined on-site after one cycle of operation were later measured in the hot cell after three or four cycles of operation. In addition, two of the Zion fuel rods were examined both on-site and in the hot cell. Of the 66 total traces, 39 traces confirmed the absence of measurable interpellet axial gaps. Of the 27 traces in I which axial gaps were measured, 6 traces indicated only one gap present in the rod, while the remaining 21 traces indicated the presence of 2 or more small gaps distributed I over the length of the rods. The maximum individual axial gap (uncorrected for differential contraction) was only 0.103 inches. Figure 3.1-1 illustrates the distribution of observed uncorrected axial gaps as a function of axial elevation. E There is a slight trend of increasing axial gap as a function of elevation, but in general, most of the gaps are
$ very small and appear to be almost randomly distributed over i
the length of the rods. Figure 3.1-2 illustrates the gap data corrected for thermal contraction effects. All but four of the observed axial gaps would be expected to be completely closed at the hot operating condition. Of the four remaining axial gaps, the largest gap size is less than [ ]* inches. *(a,b,c) 6
w www_------- -- __ . _ _ l l I t It is concluded, based on these y-scan data, that axial gaps do not typically occur in current Westinghouse fuel product. When gaps do occur, they are very small (( )+ +(a,b,c) I inches or less). Axial gaps of this magnitude do not impact local power peaking. 3.2 Flux Map Dah The Incore flux trace system provides t'other method for obtaining information on fuel densifi' <1t1 during operation. With this system the loc- nr .cron flux is measured by noveable fissf3n eknabere , ach traverse the I full length of the active tuel in instrumented locations in the core. When significant interpellet gaps are present, local flux spikes can be detected by the Incore system. In the early evaluttion of fuel densification cited here, only I flux spikes of 1% or greater were recorded, since smaller flux spikes could be caused by recorder perturbations or measurement system noise. While this system does not I provide as precise a measurement for axial gaps as the y-scanning measurement, it does allow the detection of significant axial gaps at low burnups when the maximum not densification is expected to occur. As part of early investigations of the densification phenomenon, incore flux traces were obtained from eight operating Westinghouse reactors and examined for flux spikes indicative of the presence of interpellet gaps'. Nineteen pressurized and 12 unpressurized regions were examined. I Only the data for the nineteen pressurized fuel regions were considered in thin evaluation. Most of these regions were fabricated with early densifying fuel, and are therefore not ig indicative of current fuel performance. However, some
'g regions wete fabricated with sintering temperature and/or initial density which were only slightly lower than the current Westinghouse specification. Data from these regions are therefore bounding with respect to the densification I- effects for current fuel.
I Table 3.2-1 summarizes the stack length and flux blip data which were obtained from the Incore system measurements (note that this table was reproduced from Reference 1). No as flux spikes were observed in any fuel assembly which was g fabricated with mean sintering temperature of ( )+ or +(a,c) greater. These data provide added confirmation that current stable fuel, fabricated with higher sintering
.E temperature and initial density, will nct generate 5 significant densification related axial gaps.
I 7 I
w 3.3 Other Data The densification and swelling behavior of current generation Westinghouse fuel is well understood. Fuel performance models for densification and swelling have been developed for Westinghouse fuel based on both measured fuel stack length change data and direct density measurements on I irradiated fuel. The current Westinghouse densification and swelling model, along with the data base of measured fuel density, has been reviewed by the NRC in Reference 10. I These data confirm the densification stability of current Westinghouse fuel, with maximum not fuel densification expected to occur early in life. It is also noted that a large number of on-site fuel examination campaigns have been conducted on current stable Westinghouse fuel. These examinations include visual I examination of peripheral fuel rods during refueling operations, TV visual examinations of selected fuel assemblies, and in some cases, detailed individual ftel rod examinations. It is estimated that several thousand fuel rods have been examined on-site, with no evidence of any densification related clad flattening events. It is
)* inches *(a,c)
I recognized that relatively large axial gaps (( or greater) would be required for clad flattening to occur, so that the absence of clad flattening does not address the issue of flux spiking associated with intermediate gaps. I However, this very large experience base does provide substantial assurance that the densification behavior of current generation Westinghouse fuel is well controlled. The absence of large axial gaps in the fuel column for current generation Westinghouse fuel is consistent with the current understanding of fuel densifica. ion. For current I pressurized fuel rods, maximum net fuel densification is achieved prior to when fuel-clad contact is expected to occur. The fuel stack is free to move within the cladding, and the fuel pellets therefore settle without gaps to the densified fuel stack length. It is also noted that the conclusion that current generation fuel is stable with respect to fuel densification and not susceptible to large axial column gap formation is consistent with the experience of current fuel vendors". 4.0 Conclusions Based on the above described data base, which confirms that axial pellet column gaps due to fuel densification occur very infrequently and are extremely small (essentially undetectable with Incore flux trace systems) for current fuel, it is concluded that an augmentation factor of 1.0 is 8
appropriate for the Westinghouse fuel to be supplied for the H Fort Calhoun reload cycle designs. L 5.0 References
- 1. Hellman, J. M. (Ed. ) , " Fuel Densification Experimental Results and Model for Reactor Applir:ation , " WCAP-8218-P-A I (proprietary), March 1975.
- 2. Tarby, E. J., et. al,, " Interim Report Zion Unit 1 Cycle 2 Fuel Performance," WCAP-9255, January 1979 (Non-Proprietary).
- 3. Balfour, M. G., et. al., " Zion High Burnup Fuel Hot Cell Examination Program," WCAP-10473, February, 1984 (Non-I, Proprietary).
- 4. Nayak, U. P., et. al., " Final Report EP80-16, Hot Cell I Examination of Zion Fuel Cycle (Non-Proprietary).
5," WCAP-10543, June, 198S Kunishi, H. and Nayak, U. P. , " Zion High Burnup and Contacting I 5. Rod Hot Cell Program," WCAP-10544 (Proprietary). DeStefano, J., et. al., " Interim Report Surry Unit 2 End-c - I 6. Cycle 2 Onsite Fuel Examination of 17X17 Demonstration Assemblies After One Cycle of Exposure," WCAP-8873, January, 1978 (Non-Proprietary).
- 7. Kuszyk, J. A., " Hot Cell Examination of Surry Three- and Four-Cycle 17X17 Demonstration Fuel," WCAP-10514 (COE/ET 34 014-14) ,
June, 1984 (Non-Proprietary). l 8. Balfour, M. G., et. al., "BR-3 High Burnup Rod Hot Cell Program," WCAP-10238, November, 1982 (Non-Pror"ictary).
- 9. Balfour, M. G., et. al., "Zorita Research and Development Program Final Report," Volumes 1 and 2, WCAP-10180, September, 1982 (Non-Proprietary).
- 10. Weiner, R. A., et. al., " Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August, 1988.
- 11. "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Volume 5: Evaluation of Interpellet Gap Formation and Clad Collapse in Modern PWR Fuel Rods," EPRI NP-3966-CCM, April 1985.
9 i
M M M M M M M M M M M M M M M M M M Table 3.1-1 on-Site Gap Heasurement Data from the Zion Extended Burnup Program Corrected Cumulative Number Axial Individual Individual Rod Number Rod Gap of Location Gap Gap Average Operating Program ID Site Gaps from Botton Size Size Burnup Cycles (inches) (inches) (inches) (inches) MWD /MTU Zion 687 0.147 2 54.6 0.044 0.000 16600 1 133.3 0.103 0.000 Zion 652 0.057 2 58.6 0.030 0.000 18400 1 60.1 0.027 0.000 Zion 616 0.073 2 47.6 0.047 0.000 19900 1 65.8 0.026 0.000 Zion 663 0.024 2 105.5 0.007 0.000 29980 2 133.1 0.017 0.000 Zion 691 0.041 3 12.7 0.007 0.000 30350 2 21.1 0.009 0.000 108.6 0.025 0.000 Zion 657 0.043 2 134.8 0.022 0.000 30690 2 137.3 0.C21 0.000 Zion 685 0.087 5 13.3 0.005 0.000- 30950 2 22.6 0.043 0.009 30.6 0.013 0.003 55.0 0.006 0.000 107.2 0.020 0.000 Zion 610 0.041 1 142.0 0.041 0.000 31140 2 Zion 640 0.060 5 6.6 0.010 0.000 31340 2 44.0 0.005 0.000 90.2 0.026 0.000 133.1 0.014 0.000 139.2 0.005 0.000 Zion 648 0.034 3 13.3 0.008 0.000 31630 2 14.5 0.018 0.002 60.9 0.008 0.000 Zion 677 0.023 2 6.1 0.013 0.004 31820 2 81.3 0.010 0.000 Zion 642 0.044 4 12.3 0.009 0.000 33100 2 12.9 0.007 0.000 29.8 0.005 0.000 42.3 0.023 0.000 10
y ( y--- 3 g-Table 3.1-2 Hot Cell cap Measurement Data from the Zion Extended Burnup Program Corrected Individual Individual Rod Number Cumulative Number Axial Gap Average operating Rod Gap of Location Gap Size Size Burnup Cycles , Program ID Size Gaps from Bottom l (inches) (inches) { inches) (inche-
- MWD /MTU
*(b,c) !
19000 1 Zion 616 0.030 1 32900 2 Zion 642 0.040 2 37900 3 Zion 699 0.030 2 45900 4 Zion 624 0.030 2 47500 4 Zion 614 0.000 0 54400 5 Zion NIS 0.022 1 55100 5 Zion B8 0.000 0 5 0 55100 Zion B7 0.000 55200 5 Zion A7 0.059 3 55200 5 Zion A8 0.123 5 55300 5 Zion 618 0.000 0 55300 5 Zion B9 0.000 0 5 1 55400 Zion GIS 0.043 55400 5 Zion A9 0.000 0 55900 5 Zion A10 0.000 0 0.145 3 55900 5 Zion B15 55900 5 Zion CIS 0.094 3 11
M M m M M M 'W . M M M W- m m m . mm Table 3.1-3 On-Site Gap Measurement Data t' rom the Surry 17X17 Extended Burnup Program corrected Cumulative Number Axial Individual Individual Rod Number Rod Gap of Location Gap Gap Average Operating Program ID Size Gaps from Bottom Size size Burnup Cycles. (inches) { inches) (inches) (inches) MWD /MTU Surry 2 500 0.000 0' 7470 1 Surry 2 501 0.000 0 8080 1 Surry 2 502 0.000 0 9090. 1 Surry 2 503 0.000- O 6310 1 Surry 2 505 0.000 0 7470 1 Surry 2 506 0.000 0 7620 1 Surry 2 508 0.000 0 8290 1 Surry 2 509 0.000 0 8080 1 Surry 2 510 0.000 0 9130 1 Surry 2 511 0.000 0 7990 1 Surry 2 512 0.000 0 9090 1 Surry 2 514 0.000 0 7990 1 Surry 2 515 0.000 0 6310 1 Surry 1 105 0.000 0 7940 1 Surry 1 106 0.000 0 8000 1 Surry 1 108 0.097 1 64.8 0.097 0.000 8000 1 Surry 1 113 0.119 2 54.7 0.054 0.000 7800 1 89.0 0.065 0.000 12
y M J p D FOA T i J J Table 3.1-4 Hot Cell Gap Measurement Data from the Surry 17X17 Extended Burnup Program corrected Individual Individual Rod Number ! Cumulative Number Axial Average Operating Rod Gap of Location Gap Gap l Gaps from Bottom Size Size Burnup Cycles Program ID Size (inches) { inches) (inches) { inches) MWD /MTU 0 44300 4 Surry 501 0.000 43800 4 Surry 502 0.000 0 29600 3 Surry 507 0.000 0 44500 4 Surry 509 0.000 0 44300 4 Surry 511 0.000 0 512 0.000 0 43900 4 Surry 43400 4 Surry 513 0.000 0 30000 3 Surry 515 0.000 0 39600 4 Surry 888 E9 0.000 0 40000 4 Surry 888 M13 0.000 0 28900 3 Surry 889 D7 0.000 0 30000 3 Surry 889 M9 0.000 0 l 13
- I I Table 3.1-5 Hot cell Gap Measurement Data from the BR-3 Extended Burnup Program Cumulative Number Rod Number I Program Rod ID Gap Size Gaps of Average Burnup Operating Cycles (inches) MWD /MTU BR-3 24I6 0.012 >l 60100 3 BR-3 28I6 0 53300 3 I BR-3 BR-3 BR-3 30I8 36I8 11115 0.008 0
0
>l 59800 61500 48600 3
3 2 I I I I I I I-I I 14 I
I I Table 3.1-6 Gap Measurement Data from the Zorita High Burnup Program Cumulative Number Rod Number I Program Rod ID Gap Size of Gaps Average Burnup Operating Cycles (inches) MWD /MTU Zorita 285 0.000 0 29600 1 Zorita 244 0.008 1 38900 2 Zorita 235 0.000 0 49200 3 I I I I I I I I I I I
;I 15 I
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1 W W W W W m M M m- M M M M M M m m I I 1 ' Table 3.2-1 Incore Fuel Stack Height Data From 19 Pre-Pressurized Fuel Regions No. of Average No. Geo . Density Sintering Stack Length Burnup at Assemblies of Blips Decrease at Maximum Examined Observed of Fabricated Temperature per Max. (%) for stack Region Fuel (*C) Length Assembly Min. Mean Avg Max (GWD/MTU) Number (% T.D.} + +
+
10.0 6 1 92.0 1 93.9 5.0 2 3 93.9 5.0 3 4 90.9 11.0 4 6 91.9 9.0 5 8.0 2 6 90.9 2 93.7 5.0 7 5.0 5 8 92.9 4 91.9 5.0 9 '4. 7 3 10 93.8 4 92.8 ?.7 11 5.0 6 12 92.1 7 91.8 5.0 13 10.0 7 14 90.8 9.0 2 15 91.8 8 90.8 9.0 16 10.0 4 17 92.2 8 92.7 6.0 18 6.0 4 19 93.0 -
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