ML20002C538

Densification Data of LWR Mixed Oxide Pellets & Comparison W/U Dioxide Pellets

ML20002C538
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	01/31/1975
From:	Galbraith K, Gerrald L, Rasmey Robinson SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20002C536	List: ML20002C535 ML20002C538
References
XN-75-011(NP), XN-75-11(NP), NUDOCS 8101100434
Download: ML20002C538 (26)

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XN-75-11 NP g 8 0y Mbb JANUARY 1975 Amtlpf.WM'hY [Id i

DENSIFICATION DATA i

0F LIGHT WATER REACTOR MIXED OXIDE PELLETS AND COMPARISON WITH UO PELLETS 2

Prepared by:

L. D. Gerrald K. P. Galbraith

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i XN-75-11 Np January 15, 1975 DENSIFICATION DATA 0F LIGHT WATER REACTOR MIXED OXIDE PELLETS AND COMPARISON WITH UO PELLETS 2

Prepared by L. D. Gerrald K. P. Galbraith Approved:

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G. A. Sofer, hanager Fuel Design and Engineering

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R. K. Robinson, Manager Process Engineering Approved:

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I 2i/73' W. S. Nechodom, Manager Licensing and Compliance

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EQON NUCLEAR COMPANY,Inc.

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i TABLE OF CONTENTS Page 1

.l.0 INTRODUCTION..............-............

4 1

2.0

SUMMARY

1 3.0 DATA SOURCES 3

4.0 DENSIFICATION DATA

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4.1 Experimental Densification Data 7'

4.2 Discussi on of Da ta....................

5.0 PREDICTION OF MIXED OXIDE DENSIFICATION WITH THE GAPEX DENSIFICATION MODEL..................-.

9 5.1 Comparison of Measured Density Change with 10 Predicted Density Change..-...............

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5.2 Comparison of Measured to Predicted Rate of Densification 10 4

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6.0 CONCLUSION

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REFERENCES.............................

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APPENDIX A - EEI Densification Data APPENDIX B - Discussion of Determination of Fuel Density Change from In-Reactor Pellet Stack Height Measurements....

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P LIST OF FIGURES Page 1

Apparent Density Change for Irradiated Fuel versus Exposure....................

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2 Comparison of Fuel Density Changed After Irradiation and Resintering.................

6 4

3 Comparison of Measured and Predicted Density Change Due to Densification.................

11 4

Comparison of Measured and Predicted Rate of Densification.

12 LIST OF TABLES 1

Big Rock Point In-Core Densification Data for 00 and M0 Fuels....................

5 2

B-1 Irradiated Pellet Stack Height Measurement for BRP Bundle D-72 19 i

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XN-75-11 NP I

l.0 IN1RODUCTION Data on in-reactor densification of mixed oxide (MG) and U0 fuel are 2

presented. Also data on the densification of both fuels during out of core thermal resintering are presented and compared with the in-react densifi-cation. These comparisons are made to ascertain the conservatism of appli-cation of the GAPEX model to light water reactor mixed oxide pellet type fuel.

2.0 SUMMRY The relative relationship between in-core densification of UO and 2

j mixed oxide (MO) fuel as well as a comparison of in-core and ex-core (resintering) data indicate that the M0 fuel exhibits density changes while UO density 2

i The comparison of in-core M0 densification data with predictions made using the GAPEX densification model* showed the prediction model to be in both the magnitude and rate of densification, and in terms of j

gap coefficients calculated by this model.

3.0 DATA SOURCES Data on U0 and M0 densification were derived from pellet stack 2

height measurements on Exxon Nuclear Company fuel in the Big Rock Point rea ctor. Additional data on U0 densification were taken from the EEI/EPRI 2

• XN-73-25 -Revision 0, updated as given in XN-74-53 and XN-209 -

Supplement 4 - Revision 1.

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fuel densification program.

The latter data were excerpted from EEI/EPRI Progress Report / December 1974. A description of the EEI program is pro-vided in Appoldix A.

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4.0 DENSIFICATION DATA Tne relative relationship between in-core densification of U0 and 2

mixed oxide (MO) fuel as well as a comparison of in-core densification and ex-core resintering data is presented.

4.1 Experimental Densification Data Figure 1 provides a comparison of the relative magnitude of apparent fuel densification between the M0 and 00 fuels measured by Exxon 2

Nuclear Company. The figure is a plot of percent density change,aoIRR,versus exposure.

The apparent density change,apIRR,is calculated from fuel stack height measurements on fuel irradiated in the Big Rock Point Reactor.

f Figure 2 is a plot of ap!RR (the change 1:. apparent density observed during irradiation) versus apRS (the change in density observed during resintering at 1700'C). The figure shows the maximum irradiated density change observed for each of the nine EEI base program fuel types together with 17 Exxon Nuclear fuel rods as a function of the density change observed after resintering. The Exxon Nuclear data plotted on Figure 2 i

consists of ten MO observations and eight U0 bservations.

The Exxon 2

Nuclear observed density changes resulting from resintering for both the M0 and U0 fuel were based on a resintering test at 1700'C for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The 2

density changes resulting from in-reactor densification were derived from gamna scan, eddy current and plenum gage measurements on pellet stack height after exposure-in the Big Rock Point Reactor assuming isotropic densification.

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Apparent Density Change for Irradiated Fuel, ApIRR, %TD o

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XN-75-11 NP TABLE I BIG ROCK PolNT

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IN CORE DEN 51FICAil0N DATA TOR U0 AND M0 FUELS 2

MIRR,1 D Fuel Bundle E qolure.

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Seq.

Bundle Rod No.

No.

Identification 1973*

1974*

Type 1973*

1974 me

• Date esamined.

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Resintering Density Change, ApRS, %TD FIGURE 2 Comparison of Fuel Density Change l [

Af ter Irradiation and Resintering l

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4.2 Discussion of Data

.The ' data presented in Figure 2 indicate that for both the EEI and

'the Exxon Nuclear observations, the maximum in-reactor densification is by densification resulting from a thermal resintering test conducted at 1700 C.

The data presented in Figure 2 also indicate that This behavior suggests that M0 fuel tends. to while the U0 fuel microstructural 2

form i

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Some of the fuel rods listed in Table 1 and displayed in Figures l i

and 2 were fabricated as densification test rods at Exxon Nuclear to test' methods of controlling fuel densification.

These rods were operated over a short irradiation cycle lasting 25 full power days (600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />) correspon-j ding to a bundle average exposure of 674 to 735 MWD /MIM.

The U0 short cycle 2

test rods correspor.d to numbers 54 through 57 and 59 through 62.

Rods 56 and 61 were fabricated with fuel that and thus they were expected to exhibit the of density change among the test group. The other six rods were fabricated with fuel that was and consequently were not expected to exhibit density changes.

Rods 56 and 61 did exhibit the density change (based on both

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irradiation and resintering measurements) while the remaining rods exhibited density changes.

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Theshohtcycletest.rodsincludedamixedoxidefuelrod(number 58). This rod was manufactured without.taking any special measures to

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- fuel densification. ThelM0 densification' performance obtained with this rod was 4

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t 5.0 PREDICTION OF MIXED OXIDE DENSIFICATION WITH THE GAPEX DENSIFICATION MODEL The GAPEX(1,2,3) computer program which is used to predict pellet-to-cladding heat transfer coefficients contains a densification model that was based on UO densification data.

The GAPEX densification model gives a 2

prediction of the rate of densification and the maximum change in density.

The rate of fuel densification is calculated from the following:

ap /ap

= [0.007 t]

t

<20 p

max ap /ap

= [0.2198 in(t) - 0.5134]

20

<t -

<1000 p

max Ap /ap

= 1. 0 t>

1000 p

max where:

t = effective full power hours Ap = predicted density change p

op

= maximum fuel density upon completion max of densification.

The densification rate expressions were developed from the experimental dataofHanevik,et.al.,(4) The maximum change in density, apex ' iS calculated from the following:

ap,,x

= 96.5 - (p - 2a) j where:

p = nominal as-fabricated fuel density g

a = standard deviation in the measured probability

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distribution for the fuel density.

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5.1 Comparison of Measured Density Change with Predicted Density Change Figure 3 provides a comparison of the measured density change to -

GAPEX predicted density change for Exxon Nuclear fuel irradiated at Big Rock Point. The measured density change is based on' fuel stack height measurements and reported in columns (4) and (5) in Table 1.

With the exception of three data points,all U0, the change in fuel density due to 2

densification is The three-UO2 points that fall were for The M0 densification data fall 5.2 Comparison of Measured to Predicted Rate of'Densification Figure 4 provides a comparison of the measured to predicted 4

density change versus exposure..The measured density change is based on fuel stack height measurements. This. comparison of the measured and pre-dir

.esity changes with exposure shows the GAPEX Jensification model to be 1

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GAPEX Predicted Density Change, Ap "TD p,

FIGUP.E 3 Comparison of Measured and Predicted Density Change Due to Densification

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(All data are ENC Big Rock Point Irradiated Rods Detailed in Table 1)I m _

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J Apparent Deiisity Change. ApIRR Predicted Density Change, Ap P

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6.0 CONCLUSION

The data -presented in Figures 1, 2, 3 and 4 indicate that:

-(l) Mixed oxide fuel exhibits ~

while UO densi ty ' changes -.

2 (2) -Thermal resintering tests provide I

(3) The GAPEX model.can be i

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REFERENCES

...1. - GAPEX: - A Computcr Program for Predicting Pellet-to-CladEng Hcat Transfer Coef Ycients, August, 1973, XN-73-25.

. A Computer Program for Predicting Pcllet-to-Cladding Her 2.

GAP 5X:

Transfer Coefficients, CAPEX Code Operating Manual, December, 1974,

.XN-74-58.

3.

Densification Effects on Exxon Nuclear Preocurized Water Reactor Fuct, December, 1974, XN-209 - Supplement 4, Revision 1.

'4.

Hanevik, et.al., In-Reactor Measuremente of Fuct Stack Shortening, Paper #80, presented at BNES (19?3).

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lj APPEND:X-A' EEI DENSIFICATION DATA The EEI/EPRI fuel densification program was a joint venture among EEI/EPRI and several fuel vendors. The-experimental program tested the effect

'of pellet' characteristics (pore size-distribution, grain size,. initial density, 2 powder resinterability) and irradiation parameters (fission rate', exposure, UO pressurization, axial restraint) on the densification observed during irradia-tion in GETR.

l The pellets used m the test were nine " base program" types and several

" participant" types. Data presented here is only fr.m the base program types.

These base types were fabricated at Battelle-Northwest to meet nine separate specifications on pore size distribution, grain size, etc. and then character-ized for microstructure using optical and scanning electron microscopy.

This experimental investigation generated a considerable amount of. data.

For clarity, only maximum observed in-reactor densification for each of the nine types has been included in this report. The data has been excerpted directly from a paper authorized by the program sponsors.

There have been no statistical tests of significance applied to the data.

The samples fron; eich of the nine types were resintered at various temper-atures including 1700 C fcr times ranging from 4 to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The eel data shown in this report were resintered at 1700 C for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

I Initial densities were measured at Battelle-Northwest using water immersion.

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Irradiated densities were measured at Battelle Columbus Laboratories using-

- Hg-immersion. A bias. correction factor was applied to the two measurements

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to produce the p~ublished~ ao IRR' The comparison of in-reactor densification with thermal resintering r

- (1700 C/48 hr.) densification indicates that the resintering test is a: good predictor of maximum in-reactor densification. This is' illustrated in Figure 2.

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APPENDIX B OISCUSSION OF DETERMINATION OF FUEL DENSITY CHANGE FROM IN-REACTOR PELLET STACK HEIGHT MEASUREMENTS The in-reactor densification of Exxon Nuclear fuel is derived from measurements of pellet stack height in reactor and as-fabricated for the inspected rods.

Thh assumptions required for. the derivation are as follows:

1) The fuel densification behavior is isotropic.

'2) The effects of pellet-pellet. gaps and pellet chipping are negligible.

3) The pellet stack weight is constant.

The derivation is as follows.

1). Density (p) 2 MASS (M)/ VOLUME (V) 2 V = { D L for aellet of diameter D, length L.

2)

3) Subscript 2 denotes the value of a parameter at the time o' inspection (irradiated); subscript 1 denotes the as-fabricued condition.

M /V MY 2 2 21

4) P /# 1

• M)/V)

Q 2

5) By assumption (3) M2*Mi D) L)

Therefore, p2/P1

• Y /Y2*D2 '2 l

D)

L

= q)

6) Byassumption(1)

Therefore, p2/P1 " ('l/l )

2 It follows from assumption (2) that the change in L2 from L) is a result of ap.

Therefore, the in-reactor density (p2) can be calculated from values

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of p), L, and L. The former two variables, p) and L), are /available from j

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. fabrication records. The value of L is estimated from-indirect measurements-2 of pellet stack' height af ter: gieriods of irradiation.

Three methods have been used for stack height measurements:

1) Gama ' Scan

-2) Eddy current

3) Plenum Gage Details of the length measurement methods are not presented here.

A given rod at a'given time may have been inspected using one, two or-all of the three methods.

Gama Scan is usually the method used if _ only one measurement _is taken. The precisions of the three methods are not necessarily

. equal and a bias between the methods may exist.

The An's listed in Table 1 and those plotted ~on Figure 1 are derived from the weighted average values of stack height measurements taken by the methods employed. If _only one value (method) was available for a group of rods at a

.4 given time, that was the value used. Weighted averages were adjusted to be consistent with gansna scan measurements if a single method was employed. The weighting technique is illustrated below using data from nine rods from Bio Rock Point bundle D-72.

(See Table II-1) i

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TABLE B-1 IRRADIATED PELLET STACK HEIGHT HEASb?EME:lT F0,R,BRP BUNDLE D-72 Weighted.

Average 1974L 1974 1974 (Bias-Gamma Scan Eddy Current

~ Plenum Gage Corrected)

Stack Height Stack Height

. Stack Height Stack Height Rod Number Inches Inches Inches Inches DB2-0006 67.09 67.15 67.05 67.09' 0B3-0012 67.75 67.81

~67.60

.67.71 DC4-0004 67.74 67.80 67.61 67.71 004-0005.

67.51 67.63 67.42 67.54 0D4-0007-67.62-67.87 67.50 67.67 DD4-0008 67.53 67.63 67.42 67.54~

. 1 del-0002.

67.33 67.33 67.23 67.29 DJ5-0001-67.64 67.90.

67.80 67.79 DKS-0014 67.83 67.84 67.70 67.78 M EAN --

67.568 67.662 67.481 67.568 STD DEV 0.2308 0.2615 0.2337 0.2390

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The data from the three sources are weighted on the basis of an estimate This

-of the measurement error variance for each of the three methods.

measurement error variance is estimated using a technique described in Reference 1.

The weight of each measurement method is calculated using the equation:

2

_Xi Wg=

3 2

E (%$ )

i=1 where: Si is the estimate of the measurement error variance for method i.

2 The weighted average stack height for a rod is then 3

'W ) (Stack Height)j ackHegb" i

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i=1 Since the gamma scan data is used in the absence of other data, the weighted average values are corrected for a bias between the gamma scan and weighted average values.

Bias I (Average Stack Height)GS - (Average Stack Height)gt,4yg, Since the best estimate of the measurement error variances for the three methods is obtained when comparing all three simultaneously, the weights for two methods is calculated using the three method results. The bias is computed from all observations using the two methods.

The calculated measurement error variances and corresponding weights are listed below.

1.

Frank E. Grubbs, Am. Statistical A:sn. J, V.43,1943, pp 243-264.

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ALL THREE. METHODS 5.72(10'3) 0.245.

0.076-Gamma Scan.

3.43(10-3) 0.408 0'.059 Eddy' Current.

4.03(10-3).

0'.347 0.063' Plenum Gage Bias = - 0.003"

11. GAMMA SCAN AND PLENUM GAGE 2

5 W

S 5.72(10-3) 0.413 0.076 Gamma Scan

. 03(10-3) 0.587 0.063 4

Plenum Gage Bias = O.091" The data resulting from these adjustments are used in the 'densification.

calculations.

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