GEAP-13603 AEC RESEARCH AND DEVELOPMENTREPQRT NOVEMBER 1970

CENTER TEMPERATURE MEASIKEIENTS &4;rqElJ

OF MIXED OXIDE FUEL ZERO TO 3,000 MWd/Te

0. W. SANDUSKY A. E. CONTI

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER

Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

L E G A L N O T I C E This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, expie-, o r implied, o r assumes any legal liability or responsibility for the accuracy, com- pleteness or usefulness of any information, apparatus, product or process disclosed, o r represents that its use would not infringe privately owned rights.

4 *:t,'.,;> t :J ,

G EAP- 13603 AEC Research and

Development Report ~November 1970

CENTER TEMPERATURE MEASUREMENTS OF MIXED OXIDE FUEL FROM ZERO TO 3,000 MWdITe

R . R. Asamoto P. E. Bohaboy D. W. Sandusky

A. E. Conti

Approved:

Fast Ceramic Reactor Development Program ' Development Engineering

Prepared for the United States Atomic Energy Commission

Under Contract No. AT(04-3)-189, Project Agreement No. 10

Printed in U.S.A. Available from the clearing House for Federal Scientific and Technical Information Naliunal Bureau 6 / Star~dar* U.S. Deparrmenr of Commerce

Springfield, Virginia Price: $3.00 per copy

-

BREEDER REACTOR DEVELOPMENT OPERATION ~ G E N E R A L ELECTRIC COMPANY SUNNYVALE. CALIFORNIA 94086

LEGAL NOTICE

T h i s report was prepared as a n account of Government sponsored work . Neither t h e L1nited States, nor the Com~taission, nor any person acting orr behalf of t he Commission: A. Makes any warranty or representation, expressed or implied,

w i t h respect t-o the accuracy, completeness, or u.refulness of t h e in format ion co?i;ained in this report, or that t h ~ ' use of any information, appuratus, method, or process disclosed i n this report ?nay not infringe privately owned rights; or '

. . . . B. Assumes any liabilities w i t h respect to the use of , or for dam. : .

ages resulting from the use of a7ry infor;)ua/ion, apparatus, method, or process disclosed i n this report.

As used in the above, "person acting o n behalf of the Commission" includes a n j . enzployee or contractor'of the Conzmission, or em- ployee of strch contractor, t o the extent that srrch employee or contractor of t he Commission, or employee of such contractor prepares, disseminates, or provides access to , any itaformotion pursuant t o his employment or contract w i th the Con~?nission, or his employment w i th such corrtractor.

TABLE OF CONTENTS

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : . . . . . . vii : i

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 . INTRODUCTION 1

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 . EXPERIMENT DESIGN 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 Fuel Pin Design 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2 Capsule Design 2

. . . . . . . . . . . . . . . . . . . . . 3.3 Irradiation Facility and Operational History 2

4 . TEST RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1 Linear Power Calculations 9

. . . . . . . . . . . . . . . . . . . . . . . . . . 4.2 Fuel Clad Surface Temperatures 11 4.3 Fuel Integral Conductivity and Fuel-Cladding Gap Coefficients . . . . . . . . . . . . . . 14

. . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4 Post-Irradiation Examination , 14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.5 Metallography 20

. . . . . . . . . . . . . . . . . . . . . 4.6 Post-Irradiation Thermocouple Calibration 26

i /.I 5 . DISCUSSION OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . REFERENCES 32

APPENDIX

A . TUNGSTEN.RHENIUMTHERMOC0UPLES . . . . . . . . . . . . . . . . . . . . . . . 33

B . FUEL PELLET PARAMETERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34

. . . C . THERMAL CONDUCTIVITY CURVES FOR 304 SS. 316 SS. 321 SS. Zircaloy-2 AND NaK 36

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D . GAMMA HEATING 36

E . DETERMINATION OF FUEL CLAD SURFACE TEMPERATURE . . . . . . . . . . . . . 37

. . . . . . . . . . . . . . . . . . . . . . . . . . . . F . FLUX DEPRESSION FOR EIAA 42

ACKNOWLEDGMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . DISTRIBUTION . . . ; 48

G EAP- 1 3603

LIST OF ILLUSTRATIONS

Figure Title . Page

. . . . . . . . . . . . . . . . . . . . . . . . . . . . ElAA Fuel Pin Drawing 3 . . . . . . . . . . . . . . . . . Typical Mixed Oxide Pellets Used in Capsule E l AA 4

. . . . . . . . . . . . . . . . . . . . . . Radial Cross-Section of Capsule E 1 AA 5 1/16 lnch and 20 Mil 0.d. Chromel-Alumel Thermocouples Around Fuel Pin Cladding . . 6

. . . . . . . . . . . . . . . Irradiation History of Capsule E l AA-GETR Cycle 97 7

. . . . . . . . . . . . . . . Irradiation History of Capsule ElAA-GETR Cycle 98 8 . . . . . . . . . . . . . . . . . . . Interfaces between Thermocouples No 4 and 6 10

. . . . . . . . . . . . . . Linear Power Versus Fuel Temperature (Center) and Time 12 Increase in Fuel Center Temperature During Constant Linear Power Operation . . . . . 13

. . . . . . . . . . . . . Integral Conductivity Cure (Pugv2 uOs8) 02.00 Above 5 0 0 ~ ~ 15 Fuel-Cladding Gap Coefficients . . . . . . . . . . . . . . . . . . . . . . . . . 16 Fuel Integral Conductivity Values at Various Irradiation Times . . . . . . . . . . . . 17 Capsule E 1 AA Gross Gamma Scan . . . . . . . . . . . . . . . . . . . . . . . . 18 Post-Irradiation Contact of Chromel-Alumel Thermocouple (No . 18) with Fuel Pin Cladding 19 Fuel Pin E l AA Sectioning Diagram . . . . . . . . . . . . . . . . . . . . . . . 21 Autoradiographs of E lAA Fuel Pin . . . . . . . . . . . . . . . . . . . . . . . 22 Diametral Composites of Central Fuel Zones . . . . . . . . . . . . . . . . . . . 23 Vost-Irradiation Fuel Annulus of Section "F" . . . . . . . . . . . . . . . . . . . . 24 Fuel Cladding Interaction . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 Cladding in the Etched Condition . . . . . . . . . . . . . . . . . . . . . . . . 25 Calibration Data of Irradiated W5ReIW26Re Thermocouples . . . . . . . . . . . . . 27

. . . . . . Relationship Between Fuel Cladding Surface.'Temperature and Linear Power ; : 29 Temperature Profiles Across Fuel 0per1ating at 15 kW/ft . . . . , . . . . . . . . . . 31 Relation of Outer Thermocouples Next to Fuel Cladding . . . . . . . . . . . . . . . 38 Relation of Outer Thermocouples to a 20 Mil Thermocouple Next to the Fuel Cladding . . 39 Cladding Surface Temperature Determined Using a 20 Mil and 62 Mil Thermocouple . . 40 Cladding Temperature Related to Thermocouple Temperature . . . . . . . . . . . . . 41

. E l AA Self Shielding Factor R = 0:1095 Inch . . . . . . . . . . . . . . . . . . . . 45 Power Shape Factor for E l AA . . . . . . . . . . . . . . . . . . . . . . . . . 46

LIST OF TABLES

Table Title ' Page

3- 1 Fuel Pin Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 Fission Gas Analysis 20

A- 1 Typical Chemical Analysis of Ultra-High-Puiity (99.9 Percent) Be0 . . . . . . . . . . 33

A-2 Calibration Data for Tungsten-Rhenium Thermocouples . . . . . . . . . . . . . . 33

0- 1 Isotopic Distribution of Fuel Batches and Stoichiometry of Sintered Pellets Utilized in . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Capsule EIAA . . : 34

8-2 Specific Parameters of Pellets in F U ~ I Pin E l AA . . . . . . . . . . . . . . . . . . 34

8-3 Fuel Column Length. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35

E- 1 Comparison of Fuel Clad Temperatures Predicted by.Calculation and a Geometric . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Approach. 42

ABSTRACT

The center temperature and the power output of an encapsulated plutonium-uranium

oxide fuel pin was directly measured during irradiation in GETR. Linear powen up to

75 k WRt and central temperatures to 2 7 0 0 ~ ~ were measured during the experiment

which extended to 3000 MWd/Te.

The experimental results indicate (7) a decrease in the effective thermal conductivity when the fuel pin is thermally cycled at center temperatures below 1650°C, and (2) a

gradual decrease in the heat transfer properties with increasing fuel burnups to

3000 MWd/Te. The first effect is attributed to irradiation damage and thermally induced

cracking in the fuel which decreases the effective heat flow. This decrease disappears

when the central temperature is increased to temperatures where fuel restructuring is

possible. The second effect resulted in increased fuel center temperatures of 100 to

2 0 0 ' ~ when compared to fuel with essentially zero burnup. This relationship was noted

for all linear powers to 75 k W/ft'and for fuel center temperatures to 2 1 0 0 ~ ~ .

1. INTRODUCTION

Mixed plutonium-uranium oxides are a prime candi- date fuel for the liquid metal fast breeder reactor. Their use requires a thorough knowledge of the material's physical.

properties, including its heat transfer characteristics. The ,di-

rect measurement of these characteristics during operation of the fuel i s difficult for three reasons. First, temperatures inside a fuel pin are difficult to measure, and thermocouple calibration is affected by radiation. Second, the heat de- livered by each part of the fuel is influenced by the shielding of the other parts. Third, the temperature drop across the fuel-cladding gap cannot be determined to a high precision.

The purpose of thjs experiment was to obtain direct measurements of the thermal performance o f ( P U ~ . ~ U0.8 )01 . 9 , fuel in a typical reactor environment and to determine any changes in properties with linear

power and burnup. Numerous attempts have been made to measure

indirectly the effective thermal conductivity of oxide fuel

pins. However, only a few experiments have been success- ful in obtaining a direct reading by placing temperature sensing devices in the fuel column. Christensen and ~ l l i o ' . used tungsten-rhenium thermocouples located at various radial fuel positions to measure the temperature distribu- tion in irradiated U02 up to center temperatures of 2150°C. Coplin, et al.' used a gas bulb thermometer to

monitor U02 fuel center temperatures to 2 7 5 0 ~ ~ . In an earlier experiment under this proyrarn, the center fuel temperature of a mixed plutonium-uranium oxide fuel pin in a thermal flux environment was mo,nitored directly with both a tungsten-rhenium thermocouple and a gas bulb therm~rneter.~ No significant conductivity change was noted during that 400 hour irradiation. The present experi- ment attempted to improve on the previously reported data by monitoring more precisely the fuel pin linear power and comparing this parameter with the corresponding fuel center temperature.

2. SUMMARY

The center temperature and the power output of an thermal conditions, annealing of irradiation damage in the encapsulated plutonium-uranium oxide fuel pin was , fuel and closing of cracks due to fuel thermal expansion directly measured during irradiation in GETR. Linear occurred. powers up to 15 kW/ft and central temperatures to 2 1 0 0 ~ ~ were measuted during the experinie~~l wll icl~ extended to 3000 MWdITe.

The thermal conductivity of the fuel, the cladding surface temperatures, and operating linear powers were all calculated from temperature data obtained during irradi- ation. Tungsten-rhenium thermocouples were used to measure fuel ccntcr tcmpcratures to 2 1 0 0 ~ ~ , while chromel-alumel thermocouples were employed to obtain temperature dera at various radial atid axidl positions out- side the fuel pin.

For the first 120 hours of thermal flux Irradiation, fuel center temperatures were maintained below 1650°C. The experimental data indicated that the effective thermal conductivity of the (PuOe2 Uo .a ), . 9 7 fuel decreased during

this pcriod. This decrease was attributed to: (1) irradiation damage to the fuel (effect on conductivity is greatest below 1000°C); and (2) the formation of cracks in the fuel result-

With increasing, fuel burnup (up to -3000 MWdITe, 280 hours of irradiation), a gradual rise in the fuel cknter temperature from 2000" to 2 1 0 0 ~ ~ was noted, while oper- ating at a constant power of 15 kWIft, Similar increases of 100 to 2 0 0 ~ ~ in the center temperature were reqiired to maintain any linear power betyeen 7 and 15 kW/ft. as pro- duced by that fuel at essentially zero burnup. Assuming that the fuel-gap conductivity remained constant a l u ~ ~ y with no thermocoukle decalibration, the factors contri-

buting to the increased center .temperature include tuel stoichiometry shifts, fission product buildup, and forma- tion of a radial plutonium gradient within the fuel. Densifi- cation of the fuel would increase the thermal conductivity. A quantitative evaluation of the above phenomena i s impossible because of the lack of sufficient thermal con-

ductivity data on plutonium-uranium oxides, and the large

uncertainties involved with the assumption made.

ing from thermal stresses. The effective decrease in heat A post-irradiation check of irradiated tungsten- transfer within the fuel pin resulted in fuel center tempera- rhenium thermocouples indicated a reduction in emf

tures increasing an average of 100°C from initial values for output with temperature. This apparent decalibration

a given linear power. This decrease. in heat transfer charac- suggests that the increase in the fuel center temperature teristic was apparently eliminated or minimized when the with burnup may have been slightly greater than the data fuel temperature was increased above 1650°C. Under these indicated.

3. EXPERIMENT DESIGN

3.1 FUEL PIN DESIGN The 316 stainless steel fuel pin cladding was

0.250 inches in diameter by 9.5 inches long, and contained

annular (Puo.2 UO.* )ol . 9 7 fuel, Figure 3-1. Three 114-inch long U02 pellets were located at both the top and bottom of the 6-inch column of fuel to provide insulation. These insulator pellets were sandwiched between 114-inch long

stainless steel dummy pellets, which provided a guide for inserting the tungsten-rhenium thermocouples. A 1-inch long fission gas plenum was designed at the top of the fuel pin, and contained a stainless steel retainer spring (spring constant of 3.1 Iblin.). The spring was compressed approxi- mately 1/16 inch and so exerted a force of about 0.2 Ib, or a stress of about 5 psi on the fuel. Two 1116-inch o.d., W5RelW26Re thermocouples were inserted 2 inches into the mixed oxide fuel column, one from the top and the other from the bottom of the fuel pin. Details of the W/Re thermocouples are discussed in Appendix A. The fuel pin was sealed i n a high purity helium (99.995% He) atmosphere.

The mixed oxide powder was made by the coprecipi- tation of plutonium hydroxide and ammonium diuranate from a solution of uranium and plutonium nitrate. The plutonium-uranium oxide powder was pre-pressed at about 35 tsi, granulated, and pressed into annular pellets at about 10 tsi. The pellets were then sintered for 2-112 hours at 1 6 0 0 ~ ~ in a helium-6% hydrogen atmosphere. Typical fabricated pellets are shown in Figure 3-2, and pertinent

fuel and fuel pin parameters are listed in Table 3-1. Detailed fuel parameters are listed in Appendix B.

3.2 CAPSULE DESIGN The capsule design consisted of concentric stainless

steel and Zircaloy tubes located around the fuel pin, which acted as both containments and thermal barriers. Standard chromel-alumel thermocouples (1116 inch 0.d. and 20-mil 0.d.) were strapped to the outside of each containment, while the annulii formed by the tubes were filled with Nak. Figure 3-3 shows a detailed cross section of the capsule and the location of each of the thermocouples. "Sets" of thermocouples were positioned in such a manner that linear

power values could be calculated on a given radial plane

from the center of the fuel to the outer containment. (A "set" comprises the three thermocouples at one angular position, excluding the center thermocouples; e.g. 3.4, and 6.) Two sets of the chromel-alumel thermocouples were located on each of the two horizontal planes defined by the hot junction of each of the tungsten-rhenium thermo- couples.

Thermocouples were securely held in place with a stainless steel band as shown in Figure 3-4. These bands were placed near the hot junction of the thermocouples to assure good contact with the stainless steel or Zircaloy sur- face. A stainless steel insert was placed around the 20-mil thermocouples so that the band would secure them firmly

I

along with adjacent 1116 inch 0.d. thermocouples.

Six nonfunctional W5RelW26Re thermocouples were placed in aluminum tubes and positioned outside the cap- sule. Following the irradiation, these thermocouples were

'

used to measure any decalibration in the thermoelectric output caused by the accumulated neutron fluence.

3.3 IRRADIATION FACILITY AND OPERATIONAL HISTORY

Capsule ElAA was irradiated in the General Electric , Test Reactor (GETR) pool in a Mark IV-A V-Raft. The capsule was placed in a facility tube allowing both radial and vertical positioning during reactor operation. The irra- diation history of the capsule during GETR cycles 97 and 98 is shown in Figures 3.5a and 3.5b. Outputs from thermocouples No. 1 (WIRe) and No. 6 (20-mil diameter Cr/AI) and reactor power have been plotted as a function of time. During the irradiation period, the capsule experienced 1 1 reactor scrams.

The total neutron flux seen by the capsule was as follows:

Fast (>0.18 MeV) = 0.53 X loz0 neutrons/cm2 Epithermal (0.17eV < E < 0.18 MeV) = 2.1 X lo2'

neutrons/cm2 Thermal (

GEAP 13603

TOP END PLUG +

EXTENSION

FISSION GAS PLENUM

SHIM

UPPER BLANKET (U021

FUEL CLADDING (304 S. St.)

END OF UPPER THERMOCOUPLE

ANNUJAR FUEL PELLETS cue, Puo$

END OF LOWER THERMoCoUPLE

LOWER BLANKET (UO*)

BOTTOM END PLUG EXTENSION

I 48 in. REF

I

Figure 31. E l AA Fuel Pin Drawing

GEAP 13603

NEG NO. 46285

Figure 3-2. Typical Mixed Oxide Pellets Used in ~ a p & l e E I AA

GEAP 13603

U02 BLANKET \ L- MIXED OXIDE 6 IN. / '02 BLANKET

TOP

/ I - 3IN. --4 CENTER CENTEn THERMOCOUPLE NO. 1

NO. 2

I THERMOCOUPLE

AXIAL POSITION: ""-7 21N. -r AXIAL POSITION: THERMOCOUPLE NO. 1, THERMOCOUPLE NO. 2,

T.C. SETS (3.4.6,) T.C. SETS (1 1.12.18)

AND (7.8.0) AND (13, 16. '17). -

*THERMOCOUPLES NO. 1 AND NO. 2 ARE TUNGSTEN-RHENIUM THERMOELEMENTS, ALL OTHERS ARE CHROMEL-ALUMEL

Figure 3-3. Radial CrossSection of Capsule E I A A

THERMOCOUPLES*

1 AN^ 2 6 AND 7

13 AND 18

4.9.12. AND 17

3,8,11, AND 16

DISTANCE - Q T.C. TO Q FUEL

0 IN.

0.135 IN.

0.158 IN.

0.261 IN.

0.490 IN.

GEAP 13603

Figure 34. 1/16 Inch and 20 Mil 0.d. Chromel-Alumel Thermocouples Around Fuel Pin Cladding

22Oor

REACTOR REACTOR MlOCYCLE. SCRAMS

END OF K SCRAMS REFUEL W 50

CYCLE

3 - nF 40- 8 g 6 s 30- 1 1 I I J a W 0 48 96 . 192 288 384 480 576 672

(21 (4) (8) 768

g E. (0 ) 112) ~(16) (20) (24) (28) (32) IRRADIATION TIME HR (DAYS)

1800

4 -,

Figure 3-5a. Irradiation History of Capsule E lAA - GETR Cycle 97

FAILED AFTER - 290 HOURS OF IRRADIATION

-

CLADDING OUTER SURFACE r

1

w G 2 % . - 2 : 1400 - CENTRAL TUNGSTEN-RHENIUM k'z THERMOCOUPLE (NO. 1)

a

CHROMEL-ALUMEL THERMOCOUPLE (NO. 6 ) IN CONTACT WITH FUEL

8 f ir" l o o o m - . 3 w u I-

600 -

200

1-

CHROMEL-ALUMEL THERMOCOUPLE (NO. 6 ) IN CONTACT WITH THE FUEL CLADDING OUTER SURFACE*

w P fi 800 a a 2 X g " w 400 a

5 5 u I-

0 % REACTOR REACTOR SCRAM

g REACTOR END-OF E;

SCRAM MIDCYCLE REFUEL' SCRAMS CyCLE

-30 - - P a 20 -- 3 g 10 5 a 0

Figure 3-5b. Irradiation Histoy of Capsule E I A A - GETR Cycle 98

TABLE 3-1 FUEL PIN PARAMETERS

Fuel Type Annular Pellets

OIM Range Pellet Density (% T.D.) Smeared Density (% T.D.) w/% Pu02 (%) Enrichment (% U-235) Pellet Diameter (Inches) Diametral Gap (Mils) Axial Gap (Mils) Fuel Length (Inches) Nominal Annulus Size (Mils) Cladding ater rial Cladding i.d. (Inches) Cladding 0.d. (Inches)

4. TEST RESULTS . .

4.1 LINEAR POWER CALCULATIONS ' ' Where K = thermal conductivity as a function of The majority of the thermocouples operated in a temperature

satisfactory manner throughout the irradiation. Only three TI ,T2 = Temperatures at R , and R2 of the thermocouples failed to perform properly. Thermo- couples No. 2 and No. 7 " (a tungsten-rhenium and c hromel-alumel thermocouple, respectively) did not

W = linear. power R , = Radial distance from the center of the

fuel to the center of the therlnocouple

respond during-the initial reactor startup. The other central reading TI.

tungsten-rhenium thern~ocouple (No. 1) oporated for , R, = Radial distance from the center of the approxir~lately 290 hours. fuel to the center of the thermocouple

A computer code was developed to aid irl the analysis reading T2 . .

of the data. Linear powers were calculated from the therm- ocouple data using the following relationship:

Integral ,conductivity curves for each of the materials W 6' K ~ T = l o g e ( R I R Z ) (4-1) . ' in the capsule were determined trom the literature and are 277 defined b\i polynomial equations (see Appendix C). The

program assumed that the temperature rec'orded by a

thermocouple was that of the NaK coolant existing at the geometric center of the thermocouple on the assumption

that there was no convection flow. -

As an illustration of calculations used in the com- puter program, the following example shows the method

* Refer to Figure 3-3 for the location of the various thermo- for determining linear power from temperatures recorded couples in the capsule. by thermocouples 4 and 6, (see Figure 4-1).

GEAP 13603

31-

FUEL

Figure 4-1. Interfaces between Thermocouples No;.4 and 6.

The linear powers were determined through an iterative process. The iteration was initiated by assuming a power of 1 kW1ft and calculating T4 from'the measured value of T, and comparing it to the actual value of To.*

Assumed power was then increased in increments of 1 kW/ft, 0.1 k ~ l f t , and 0.01 kW1ft .until the calculated value for T4 agreed with the measured value of T4. Linear

were calculated in this manner for all combinations of the three thermocouples in each "set!' (see Section 3.2 for a definition of a "set"). The calculated values within a given thermocouple "set" generally agreed within

40.1 kW/ft. Comparison of the calculated linear power values for the various sets of thermocouples did not show any trend or bias in the data, and all of the data fell .within + 0.3 kWlft of their average value. These average .vaiues were used in analyzing the bulk of the experimental data.

Linear power values as a function of fuel center temperature were obtained by adjustirlg the radial position of the capsule in the reactor (vertical adjustments were made frequently during reactor operation to keep the cap- sule centered on the peak of the neutron flux). Radial adjustments were made during normal operation of the capsule or during start-up following a reactor scram.

.Figure 4-2 illustrates the changes which occurred in the linear power-fuel, center temperature relationship as a

function of time. During the first 48 hours of capsule operation,.the linear. power appeared to be directly related

to fuel center temperatures between 1100" and 1650°C. Additional data obtained after 3 and 5 days of operation ' (50 to 96 hours), showed a change in this relationship at fuel center temperatures less than 1600°C. The change indicated a decrease in the heat transfer capabilities of the fuel pin, so that for constant linear power operation the fuel center was approximately 80 to 100°C hotter. Data recorded at center temperatures between 1600" and 2000°C appeared to be an extension of the linear relation- ship indicated during the first 48 hours of operation.

The lower curve in Figure 4-2 represents the center temperature-linear power relationship after nearly 270 hours of irradiation. Prior to this, the fuel pin operated at a constant linear power of 15 kW/ft for nearly I 50 hours. During this period, an anomaly in the fuel center temperature was observed. The central tungsten- rhenium thermocouple indicated a gradual increase in temperature from the initial value of 2000°C. A peak temperature of 2130°C was recorded, followed by a gradual decrease to 2055°C. Figure 4-3 illustrates this occurrence. A similar case of fuel center temperature increasing from 2000 to 2100°C during constant power operation has previously been observed in U02 .' Following this constant power operation, the fuel pin was slowly cycled between 7 and 15 kWIft. The data indicated a decrease in the heat .transfer capability of the! fuel pin. Refer again to the lower curve in Figure 4-2. For all linear powers, the fuel center temperature was almost 200°C higher than those measured during the start of irradiation.

The reactor experienced i t s third scram after 278 hours of irradiation. Shortly after coming back to

power, the center W-Re thermocouple failed.

4.2 FUEL CLAD SURFACE TEMPERATURES The outside surface temperature 'of the clad was

determined from the temperature of the thermocouples strapped to the clad. This could be done if uniform heat dissipation from the fuel was assumed. It was determined that the clad surface temperature was considerably higher

than the temperature of the thermocouples strapped to it. , .

For the case of a 1116-inch thermocouple, the clad surface could be as much as 1 2 5 " ~ higher than the thermocouple

when it was reading 1 100"~. A more complete discussion of the determination of clad surface temperature may be

found in Appendix E. It was also noted that a temperaturegradient existed

around the circumference of the cladding with the hottest

side facing the core. At peak operating temperature, the

A subroutine was included in the computer program to account range was as great as 50 to 60O~. This observation was coil- for gamma heating. This is discussed in detail in Appendix D. sistent throughout the life of the capsule.

TOTAL IRRADIATION TIME (hr)

REACTOR SCRAMS AT 48 hrs, 72 hrs AND 263 hrs.

264-276 HOURS

0-48 HOURS

50-120 HOURS

FUEL CENTERLINE TEMPERATURE (OC)

Figure 42. Linear Power Vems Fuel Temperature (Center) and Time

312

THERMOCOUPLE SETS O O 0 CENTRAL TUNGSTEN-RHENIUM 2100 YSED I N MAKlhG 0 0 0 THERMOCOUPLE NO. 1

LINEAR POWER CALCULATIONS A 3.4.6 00 O 0 0 0

TOTAL IRRADIATION TIME (hr)

Figure 4-3. Increase in Fuel Center Temperature During Constant Linear Power Operation

11, 12.18 0 0 0 oQ 0 U 13.16.17 = 2050 00 0 0 0

W a 3

t CP REACTOR a 0 w SCRAM

2000- 0 %O F

19M) -

O

L

- LA o_ W

1250 a

$ m o - a U)

0 0

- CHROMEL-ALUMEL THERMOCOUPLE NO. 6

L 1150- r

+@ "%deb

a 15.5- 5 2

+ ' i : I $9 . m -

I

a p 15.0 2 !k ~ 2 1 4 5 - 3 -

96 120 1 44 1 68 192 216 240 264 288

- A n 6 4 6 I) t a *

I I I I 1 I

4.3 FUEL INTEGRAL CONDUCTIVITY AND FUEL- CLADDING GAP COEFFICIENTS

The experiment design allowed reasonably accurate

values of linear power and fuel cladding surface-tempera-

tures to be calculated. These parameters were utilized to

evaluate the fuel integral conductivity, as defined in the

following equation:

Where K = fuel thermal conductivity (w/cmOc)

TE = fuel center temperature (OC)

Ts = fuel surface temperature (OC)

W = linear power (wlcm)

PSF = power shape factor* = 0.522

The fuel center temperature was measured directly

with the tungsten-rhenium thermocouple. The fuel surface

temperature was calculated through the interrelationship of

fuel cladding outside surface temperature, corresponding

linear power, thermal conductivity of the cladding material,

and the fuel-cladding gap coefficient. The latter term is

defined as the fuel surface heat flux divided by the differ-

ence between fuel surface and cladding inside diameter

temperatures. Thus,

by assuming a fuel integral curve and using Equation 44.

The base integral curve utilized was that suggested by Baily,

et at.,' and accounts for densification changes, Figure 44.

Density corrections to 97 and 99% T.D. were used at tem- peratures of 1500° and 2000°c, respectively. Nominal

linear power values were used along with average temper-

atures of the cladding outer surface determined by the

three monitoring surface thermocouples. The results of the

calculations produced a curve showing a variation of gap

coefficient with fuel center temperature, Figure 4-5. Also

plotted on the figure are data generated in an earlier

thermal conductivity experiment, capsule E I A . ~ When it is considered that E l A had a larger initial fuel-clad gap than

ElAA and argon back-fill gas instead of helium, the agree-

ment is reasonable. The curve in Figure 4-5 for ElAA

represents gap cdnductivity during initial irradiation. This

curve was used to make subsequent calculations of fuel

surface temperature. The assumption that any change in

gap conductivity during irradiation of the capsule was

insignificant, was supported by previous results.516

In equation 4-3 the fuel center temperature (T ), G linear power (W), and the power shape factor (PSF) were

readily determined. By developing a relation for gap con-

ductivity vs. center temperature, fuel surface temperatures

(Ts) could be calculated to correlate the information nec-

essary ,tu bulld a Curve for fuel integral conductivity. Cal- culated integral values were projected from a point on the

base curve, Figure 4-4, which represented the fuel surface

OIL H = - 1 nDF Ts-T

'I.D.

iernperature ro rhe cbri'esponding fuel center temperature. (4-4) The results of this evaluation are shown in Figure 4-6. The

fuel integral conductivity curve decreased after 2 days of

reactor operation. Integral values were similar to that of the Where

base curve at fuel center temperatures greater than 1600°c. H = fuel-cladding gap coefficient (Btulhr f t Z O ~ )

QIL = linear power (Btulhr f t) After a total irradiation of 270 hours (the last 150 hours at

constant power of 15 kWIft), an overall decrease in integral DF = fuel outside diameter (ft)

conductivity was evident. Ts = fuel surface temperature ( O F )

Tcl .D. = cladding inside diameter temperature (OF) 4.4 POST-IRRADIATION EXAMINATION

Gap conductivity i s a difficult parameter to charac- On completion of irradiation testing, the capsule was

terize since it is a complex function of temperature, burn- transferred to the Radioactive Materials Laboratory (RML)

up, initial fuel-cladding gap size, fuel density, cladding for destructive examination A gross gamma scan of the

thermal expansion characteristics, and the gas within the capsule, Figure 4-7, indicated no abnormalties and that the

gap. For this experiment, it was decided to calculate fuel fuel pin was intact. A puncture was made in the capsule

surface temperatures and determine gap values to check the activity the contained gas- evi- dence of fission gas release from the fuel pin was found.

The capsule was disassembled and the fuel pin was

intact. Visual examination showed that the thermocouples This unitless factor is related to the isotopic composition of the fuel and its neutron absorption cross-sections. The Dower shaoe had remained strapped to the 'ladding. An view of

factor for ElAA was calculated to be 0.522. This calculation is the fuel pin shows a thermocouple to be in intimate contact shown in Appendix F . with the cladding, Figure 48.

TEMPERATURE (OC)

Figure 4-4. Integral Conductivity Cure (PuOa2 uOa8) 02.00 Above 500°C

Figure 4-5. Fuel-Cladding Gap Coefficients

Figure 4-6. Fuel Integral Conductivity Values at Various Irradiation Times

Figure 4-7. Capsule E l AA Gross ~arnrna.~can

GEAP 13603

NEG NO. 43837

Figure 4-8. Post-Irradiation Contact of Chromel-Alumel Thermocouple (No. 18) with Fuel Pin Cladding -

After visual inspection, the thermocouples around the cladding were removed and profilometer traces made of the fuel pin. Diameter comparison with pre-irradiation diameter measurements indicated that there was no cladding strain that could be related to the fuel-cladding interaction or any other type of irradiation induced swelling.

The fuel pin was punctured and a fission gas analysis performed. The gas constituents analyzed are listed in Table 4-1. Assuming a gas yield of 0.32 atomslfission and the number of fissions to be 3.75 X lo2' fission/cc, a fission gas release value of 14.5% was calculated. This value is consistent with the relatively low center line temperature of the fuel.

TABLE 4-1 FISSION GAS ANALYSIS

Bomb Pressure Volume of Gas

0 2

N2 H2 CO

co2 CH4 Xe Kr Ar He Gamma scan (Kr-85)

1.34 Tort 2.3 cc 2.36 + 0.03% 1 1.9 * 0.06% Not detected Not detected Not detected Trace (

BLANKET MIXED OXIDE PLENUM END PLUG AND EXTENSION

BOTTOM

BURNUP -20.000 MWdKe MAXIMUIM POWER "15 kWlh PELLET DENSITY 92.3% T.D.

DIAMETRAL GAP 1 MIL

TOP

Figure 4-9. Fuel Pin E1AA Sectioning Diagram

GEAP 13603

a NEG. NO. A227-01.02.03

SECTION B*

BETA-GAMMA RADIOGRAPH -

I 2 1 10X NEG. NO. A228-01

SECTION 'ID"*

NEG. NO. A229-01.02.03

SECTION "F"'

ALPHA RADIOGRAPH

'SEE FIGURE 4-9 FOR LOCATION

Figure 410. Autoradiographs of El AA Fuel Pin

NEG. NO. A227-04, OB, 06 SEC-ICN "C"

NE G. NO. An&OtS, 08, m

Figure 4-11. Diamsttal lCP~mpodter of Central Fuel Z~nes

GEAP 13603

NEG. NO.

Figure 4-12. Post-Irradiation Fuel Annulus of Section "F"

A22Ba

THERMOCOUPLE SHEATH 12 POST-IRRADIATION 0.f367in. ANNULUS -1 50X

NEG. NO. A228-04 500X

Figure 4-13. Fuel Cladding Interaction

NEG. NO. A227-07 250X

Figure 4-14. Cladding in the Etched Condition

4.6 P O S T - I R R A D I A T I O N THERMOCOUPLE CALIBRATION

As described in Section 3.2, six "non-functional" W5RelW26Re thermocouples were positioned outside of the test capsule to accumulate fluence in order to study transmutation-decalibration effects. These thermocouples were removed at the termination of the experiment and placed in storage in the GETR pool. The total neutron flux on the wires was assumed to be similar to that calculated for the capsule wall, and was as follows:

Thermal 1.9 X 1 020 neutrons/cm2 Epithermal 2.1 X lo2' neutrons/cm2 Fast 0.53 X 1 o2 neutrons/cm2

The irradiated thermoelements (20 mil diameter W5ReIW26Re) and alumina insulation were removed from their aluminum casings and the wires were found to be very brittle. A black, flakey deposit was observed on the in- sulators and wire, and indications were that water from the GETR pool had entered the tubing during irradiation (the thermoelements were exposed to temperatures of - 1 0 0 ~ ~ ) . The hot junctions of these thermocouples, which had been

formed by mechanical twisting and spot welding prior to the irradiation, were broken during removal from the al- uminum tubing. New hot junctions were successfully formed on two of the thermocouples by crimping a piece of 1116-inch diameter molybdenum-rhenium tubing around the wires. The segment was lea than 118-inch long with an approximate 42 mil i.d. An unirradiated W5RelW26Re thermocouple with the molybdenum-rhenium junction was tested to standardize the system.

Results from the two irradiated thermocouples are shown in Figure 4-15. Physical limitations of the test ap- paratus did not allow temperature measurements above 1 2 0 0 ~ ~ . The data indicated a definite negative de- calibration, which appeared to decrease with increasing temperature. An apparent anomaly observed at temper- atures above 1 0 0 0 ~ ~ indicated that the thermoelectric out- put of the irradiated thermocouples decreased with time. No obvious explanation for this occurrence is available. Although scatter in the data is obvious, the trend indicated is similar to that previously observed by Novak and ~samoto?

.5. DISCUSSION O F RESULTS

Many variables are involved and must be carefully

considered i n attempting t o evaluate the effective thermal

conduct iv i ty o f a plutonium-uranium ox ide fuel direct ly

in-reactor. The linear power, f o r example, is one o f t h e

more impor tant parameters and is d i f f i cu l t t o measure

accurately. Evaluation o f the data f r o m this experiment

indicated t h e many sources o f potent ial errors i n properly

evaluat ing and ut i l iz ing thermocouple data. The cir-

cumferent ial temperature gradient observed could lead t o

significant errors i n power calculations i f the thermocouples

were n o t or iented in the same angular plane (assuming d i f -

ferential temperature readings f r o m thermocouples are used

t o calculate linear powers). Even w i t h the proper or i-

entation, uncertainties o f 1 t o 2 k W / f t may b e involved un-

less gamma heating effects are properly ut i l ized.

The use o f thermocouples t o estimate cladding sur-

face temperature presents another source o f possible error.

It was shown i n this investigation that either o f the t w o

methods discussed i n Appendix E may b e used t o evaluate th is temperature. The results of these t w o methods agreed

qu i te we l l and demonstrated tha t significant error results

when it is assumed tha t a thermocouple d i rect ly against the

cladding surface is reading the cladding surface 0.d.

temperature. For example, the data indicate that a

1116-inch 0.d. thermocouple w i l l understate a cladding tem-

. perature o f 1 10do F b y about 125' F. Wi th the above factors (clad surface temperature and

gamma heating) accounted for, t h e cladding inside surface

temperature may b e reasonably calculated. The heat trans-

fer characteristics f r o m th is po int t o the geometric center

o f t h e fue l is related t o three major parameters: (1) center

temperature o f the fuel; (2) t h e fuel-cladding gap co-

eff ic ient; and (3) the thermal conduct ivi ty o f t h e fuel.

Tungsten-rhenium t'herlilocouples were selected t o

measure fue l center temperatures because o f the relatively

h i g h t e m p e r a t u r e s t o b e moni tored. Previous in-

v e s t i g a t i o n ~ ~ ' indicated tha t thermal embri t t lement o f the thermoelements and thermoelectric decalibration under an

irradiat ion environment were the main areas o f concern.

T h e post-irradiation calibration- o f the non-funct ional

tungsten-rhenium thermocouples strapped t o t h e side o f the

capsule indicated tha t significant thermoelectr ic changes d o

occur. Refer t o Section 4-6. The trend is similar t o tha t

previously reported? i.e., a negative decalibration a t lower

temperatures and approaching zero decalibration as tem-

perature increases. This t rend has also been found b y other

investigators.10 O n the other hand, studies w i t h syn-

t h e s i z e d a l l o y s o f tungsten, rhenium, and osmium

( s i m u l a t i n g transmutation) have shown positive de- calibration effects dependent o n temperature, time, and

exposure.' ' ' In order t o fu l ly analyze this occurrence, one must consider the thermal spectrum, temperature gra-

dients and compositional gradients (related t o f l u x prof i le).

The available data d o suggest, however, tha t decalibration

effects may induce errors o n the order o f 5 t o 10%.

Wi th respect t o the results o f this investigation, the

implications o f the decalibration effect o n the tungsten-

r h e n i u m thermocouple data must b e considered. I n

Figure 4-2, linear power values were plotted against center

temperature. I f the observed temperature increases were the

result o f thermocouple transmutation effects, the thermo-

couple, after 288 hours o f irradiation, wou ld have ex-

perienced a large positive decalibration. (This assumes tha t

the ini t ial data obtained during the f i rst 4 8 hours in GETR are' correct and n o t substantially influenced b y thermo-

couple decalibration effects.) The post-irradiation check o f

the tungsten-rhenium thermocouples, however, showed a

d e f i n i t e n e g a t i v e d e c a l i b r a t i o n u p t o 1 2 0 0 ~ C

Consequently, the actual increases i n the fuel center tem-

peratures may have been greater than those given i n

Figure 4-2. N o evidence o f any significant irradiation induced de-

calibration o f chromellalumel thermocouples has been

observed.' 3 " It was therefore, fel t that the cladding sur-

face temperature as a funct ion o f linear power might pro-

vide some insight in to the trends observed. Figure 5-1 illus-

trates the data taken o n the various days. The in format ion

shows that whi le the fuel center temperature is gradually

increasing, the apparent temperature increase is n o t being

sensed b y the cladding. This information, along w i t h the

thermocouple decalibration argument, suggests that a real

decrease i n the fuel conduct ivi ty and/or fuel-clad gap co-

eff ic ient is occurring.

Values f o r fuel-cladding gap coefficients are n o t

readily available. Craig, e t have recently investigated

this parameter f o r mixed oxides, w i t h special attent ion o n

ini t ial fuel-cladding gap. Their data f o r pins irradiated i n a

fast f l u x (hel ium cover gas) t o 50,000 MWdITe and at

center temperature o f about 2500°c, indicate a gap co-

eff ic ient o f 1900 B t u l h r ft2 O F fo r p i n s w i t h an ini t ial 1 m i l

gap. Projection o f the linear relationship i n Figure 4-5 t o a

fuel center . temperature o f 2500°c, suggests a gap co- eff ic ient o f 1875 B t u l h r ft2 OF.

It is d i f f i cu l t t o assess the contr ibut ion o f gap co- eff ic ient t o the increase i n the fuel center temperature. lt is

believed that the linear relationship determined as a

function of temperature, Figure 4-9, is reasonable. A relatively large decrease in gap conductivity with time is

difficult to justify on the basis of earlier experiment^,^.^ The metallographic examination, Figure 4-1 1, indicated

that an increasingly tight fuel-clad gap was maintained.

There was no evidence of any significant reaction layer or

intermediate interface that could cause a decrease in the

gap coefficient. . On the basis of this experiment, the data has sug-

gested that the fuel thermal conductivity, for fuel center

temperatures between 1000' and 2000'~. does decrease

during fuel operation. The apparent decrease in fuel con-

ductivity at center temperatures less than 1600 '~ is pri-

marily the result of thermal cracks and radiation damage of

the fuel. The fact that the higher temperature data returns

to the base integral curve, Figure 4-2, suggests the dkmage

may have annealed out. Fuel surface temperatures of ap-

proximately 700' and 9 0 0 ' ~ were calculated for cor-

responding center temperatures of 1100' and 1600'~

respectively. Experimental data' indicate that pronounced

decreases in the thermal conductivity of U02 occurs after

low temperature irradiation. The effect appears to remain

to some degree even after specimens are annealed at tem-

peratures above 1 0 0 0 ~ ~ . The extent of recovery is de-

pendent on neutron dose and the annealing temperature.

The apparent decrease in fuel conductivity measured

after 288 hours of operation, cannot be explained by ir-

radiation-damage. As previously described, in Section 3.3,

the fuel pin operated at a constant power of 15 kW/ft for

about 150 hours. A fuel center temperature of 2000 '~ to

2100°C was maintained. Temperature profiles across the

fuel* during this time period are shown in Figure 5-2, and

illustrate the relatively high surface temperature of the fuel.

A burnup effect with respect to the radial plutonium

depletion was considered and for the burnups involved, no

major contribution from this effect i s apparent.

Based on the integral curve shown in Figure 4 4 .

The data reviewed in reference 5 indicated that in- reactor measurements of the thermal conductivity of U02

between 500' and 1800'~, showed no difference from un-

irradiated values. While the properties of mixed plutonium-

uranium oxides are similiar to those of U02, one major

.difference does exist, which may contribute to the results

observed in this investigation. When hypostoichiometric

mixed oxides are placed under a large thermal gradient,

oxygen migrates to the cooler portion of the fuel. it ken' ' has demonstrated this effect in mixed plutonium-uranium

oxides having Initial oxygen-to-metal ratios between 1.935

and 1.998. At temperatures less than --1200'~. the thermal

conductivity of mixed oxides is decreased by variations

from the stoichiometric condition,' . consequently stoichiometry gradients may have contributed to the ob-

served results. Unfortunately, predicted stoichiometry

changes in this experiment would tend to reduce rather than cause it to increase.

A final factor, which must be considered with respect

to the fuel thermal conductivity i s microstructural changes. The thermal conductivity of uranium dioxide has been

shown to increase markedly with density." A model was developed which incorporated density adjustment and cor-

responding thermal conductivity values for in-pile use. In-

tegral curves adjusted for fuel densification were then

generated. The net increase in the integral conductivity of

init~ally 90% and 95% T.D. specimens was calculated to be 12% and 5.5% respectively. Based on the microstructures

exhibited in F~gure 4-19, an enhancement of approximately

3% in the fuel integral conductivity would be expected

from the above model.

No conclusive explanation for the observed center

temperature increase at constant power can be made. This

same anomaly has been observed by two other investigators

testing uranium dioxide in a similar manner.' '12 ' The ex-

perimental design, test conditions, and magnitude of the

increase were very similar in a l l cases. The anomaly may be

explained by one or a combination of the factors already

discussed.

GEAP 13603

Figure 52. Temperature Profiles Across Fuel Operating at 15 kW/ft

2200

2100

2000

1900

1800

1700

1600

1500

1400-

1300

1m-

1100

lam

900-

800

700 - 0

- I I I I I I I I I I I -

-

- -

-

-

-

EDGE OF CENTE VOID -

- -

FUEL SURFACE

- I I I I I I I I I I I I I I

0.01 0.02 0.03 0.04 0.06 0.06 0.07 0.08 0.09 0.10 0.11 0.12

FUEL RADIUS (in.)

REFERENCES

1. Christensen, J. A. and Allio, R. J., "In Pile Meas-

urement of Uranium Dioxide Fuel Temperature Distribution," A.N.S. Transactions, 8, 1, p.35, (1965).

2. Coplin, D. H., et al., The Thermal Conductivity of U02 by Direct In- Reactor Measurement, March 1968, (GEAP-5100-6).

3. Sodium Cooled Reactors, Fast Ceramic Reactor De- velopment Program, 12th Quarterly Report, July-

September 7964, November 1964, (GEAP-4723). pp. 5.1-5.4.

4. Sodium Cooled Reactors, Fast Ceramic Reactor De- velopmen t Program, 73th Quarterly Report, October 1964-January 1965, February 1965, (GEAP-4795), pp. 5.1-5.7.

5. Baily, W. E., et al., "Thermal Conductivity of Uranium-Plutonium Oxide Fuels," Nuclear Metallurgy, 13, lnternational Symposium on Plu- tonium Fuels Technology, Precedings of the 1967 Nuclear Metallurgy Symposium held in Scottsdale, Arizona, on October 4-6, 1967, American Institute of Mining, Metallurgical, and Petroleum Engineers, Inc., (1968). Craig, C. N., Hull, G. R.,and Baily, W.E., Heat Trans- fer Coefficients Between Fuel and Cladding in Oxide Fuel Rods, January 1969, (GEAP-5748). Sodium Cooled ,Reactors, Fast Ceramic Reactor De- velopmen t Program, 20 th Quarterly Report, A u g u s t - O c t o b e r 1966 , November 1966.

(GEAP-5292). Novak, P. E. and Asamoto, R. R., An Out-of-Pile

Evaluation of W-Re Thermocouple Systems for Use

to 4 7 0 0 ~ ~ in Pu02 - U02," GEAP-5166. June 1966. Novak, P. E. and Asamoto, R. R., An In-Pile Evaluation of W-Re Thermocouple Systems for Use

t o 4 7 0 0 " ~ i n SEFOR Fuel, October 1968, (GEAP-5468). Aspley, E . R ., ed ., Fast Flux Test Facility Quarterly T e c h n i c a l Report , July-September 1968, December 1968, (BNWL-917). Kuhlman, W. C., Evaluation of Thermal Neutron In- duced Errors in the W/W25Re Thermocouple, 1965, (GE-TM-65-4-41) .

12. Kuhlman, W. C., Thermocouple Research at

GE-NMPO, February 1965, ( ~ ~ - ~ ~ - 6 5 - 2 : 1 5 ) .

13. Dan, G. J., Bourassa, R. R., and Keeton, S. C., "Nuclear Radiation Dose Rate Influence on Thermo- couple Calibration," Nucl. Appl., 5, 322 (1968).

14. Tranger, b. B., Experience with Stainless steel Sh ea thed Chromel-Alumel Thermocouples in Ir- radiat ion Experiments, November 12, 1965, (ORN L-TM-1320).

15. Kelly, M. J. , Change in LMF Characteristics of Chromel/Alumel and PlatinumPlatinum-Rhodium, T I D-7586(Pt-1). High Temperature Thermometry Seminar held October 1-2, 1959, at Oak Ridge National Laboratory.

16. Thermal Conductivity of Uranium Dioxide, report of the panel on thermal conductivity of uranium dioxide held in Vienna, April 26-30, 1965, Technical Report Series No. 59, International Atomic Energy. Agency, . Vienna, (1966).

17. Aitken, E. A., "Thermal Diffusion in Closed Oxide Fuel Systems," J. Nucl. Matl., 30, pp. 62-73, (1969).

18. Van Cracynest, J. D. and Weilbacher, J., "Study of

the Thermal Conductivity of Mixed Uranium and Plutonium Oxides," J. Nucl. Matl., 26, pp. 132-136,

(1968).

19. Gibby, R. L., T l~e Effect of Oxygen Stoichiometry on the Thermal Diffusivity and Conductivity of

U O . ~ ~ P U ~ . ~ S ) , - x , January 1969, (BNWL-927). 20. Asamoto. R. R., Anselin, F. L., and Conti, A. E.,

"The Effect of Density on the Thermal Conductivity of Uranium Dioxide," J. Nucl. Matl., '29, pp. 67-81, (1969).

21. Christensen, J. A., and Allio, R. J., "In-Pile Meas- urement of Uranium Dioxide Fuel-Temperature Distribution," ANS Trans., 8, 1 ,pp. 40-41, (1965).

22. Kjaerheim, G., and Rolstad, E., In-Pile Determination of U02 Thermal Conductivity Density Effects and Gap Conductance, December 1967, (HPR80).

23. Dwork, J., Self-shielding Factors for Infinitely Long,

Hollow Cylinders, January 10, 1955, (KAPL-1262). 4

APPENDIX A

TUNGSTEN-RHENIUM THERMOCOUPLES

Two tungsten-rhenium thermocouples were fab-

ricated for use in this experiment. The thermocouples were sheathed in 1116-inch 0.d. vapor deposited tungsten-22 f 2

w/% rhenium tubing with a 10 mil wall, and insulated with

double-bore, high-fired beryllia insulation. A typical

chemical analyses of the ultra-high-purity beryllia is shown in Table A-I. The thermoelements were 5 mil diameter,

tungsten-5 w/% rhenium and tungsten-26 w/% rhenium.

The manufacturer's* calibration data for the lot of wire

used is listed in Table A-2.

Hoskins Manufacturing Company ** Continental Sensing, Inc.

TABLE A- I

TYPICAL CHEMICAL ANALYSIS OF

ULTRA-HIGH-PURITY (99.9 PERCENT) Be0

Be0 Content

Impurities (ppm)

Aluminum

t3o1.01 I Calcium

Chromium

Copper Iron

Potassium

Lithium

Magnesium

Manganese Sodium

Nickel

Silicon

Strontium

Titanium

Vanadium

Zinc

Zirconium

The thermocouples were of a grounded design and

sealed under an argon atmosphere. A stainless steel

transition piece joined the tungsten-rhenium sheath to a

316 SS sheathed compensated lead wire. The wires were

brazed within the transition piece, which was located just

outside the fuel pin (the transition piece was also the end

plug for the fuel pin). The magnesia insulated compensated

lead wire (4051426) was extended to a continuous two pen

recorder. The vendor*" verified similarities in the thermo-

electric outputs of the compensated lead wire and the

W5ReIW26Re thermocouples to 1100"~. Their results

showed deviations of 1, 2, and OF at 500°, 800°, and 1 100' F respectively.

TABLE A-2

CALIBRATION DATA FOR .

TUNGSTEN-RHENIUM THERMOCOUPLES

Temperature, OF (OC) EMF (mv)

APPENDIX B

FUEL PELLET PARAMETERS.

The mixed oxide fuel pellets used in this experiment Table 8-2. The pellets are listed in the order they were were made from two batches of powder. The PuOz con- located in the fuel pin, i.e., pellet No. 1 was at the top of tent, percent enrichment, plutonium isotopic distribution the fuel pin. The length of the fuel column was measured in and stoichiometry for each batch of sintered pellets are a v-trough prior to loading, and is compared in Table B-3 to listed in Table B-1. The specific parameters for each fuel the sum of the individual pellet measurements from pellet, and each insulator and blanket pellet are, listed in Table 8-2.

TABLE B-1 ISOTOPIC DISTRIBUTION OF FUEL BATCHES

AND STOICHIOMETRY OF SINTERED PELLETS UTILIZED IN CAPSULE ElAA

BATCH 314 20.28 90.68 8.49 0.79 0.04 10.62 . . 1.979 BATCH 284 . 19.4 90.84 8.35 0.77 0.04 9.9 '. 1.971

ANALYTICAL RESULTS (INSULATOR AND BLANKET) BATCH U-63

TABLE 8-2 SPECIFIC PARAMETERS OF PELLETS IN FUEL PIN ElAA

Pellet Batch Diameter Length Weiyht Arl~~ulus Density No. No. (Inches) (Inches) (gm) (Mils) (% T.D.)

Stainless Steel Dummy Pellet 0.2197 in. 0.d. X 0.253 in. Long* U-63 0.2127 0.2567 1.317 78 93.0

0.2129 0.2551 1.356 75 ' 95.0 U-,63 0.2128 0.2560 1.358 76 95.3 B-284 0.2189 0.2503 1.394 74 92.3

0.2190 0.2403 1.343 74 92.5 0.2190 0.2477 1.373 74 91.8 0.2189 0.2573 1.431 74 92.1 0.2189 0.2553 1.417 74 92.0 0.2189 0.2404 1.330 74 91.7 0.2189 0.2541 1.410 73 91.6 0.2189 0.2510 1.397 73 91.9 0.2189 , 0.2667 1.477 73 91.5

GEAP-.13603

Table B-2 (Continued)

Pellet ' , Batch Diameter Length Weight Annulus Density

No. No. (Inches) .(Inches) (grn) (Mils) ' (% T.D.)

0.2190 0.2463 1.369 73 91.7 0.2189 0.2906 1.603 74 91.4 0.2190 0.2586 1.451 . 74 92.9 0.2189 0.2242 1.255 74 92.8 0.2189 0.2460 1.380 74 92.9 0.2189 0.2542 1.423 74 92.8 0.2189 0.2512 1.392 73 91.5

B-280 0.2189 0.2446 1.360 73 91.8 6-314 0.2189 0.2543 1..436 72 92.9

0.2189 0,2519 1.423 72 92.9 0.2190 0.2107 1.196 72 93.3 0.2191 0.2550 1.430 72 92.1

. . 0.2192 0.2484 1.400 ' 72 92.5 0.2189 0.2474 1.402 72 93.2

8-314 0.2189 0.2482 1.392 72 92.2

U-63 0.2130 0.2561 1.376 73 95.2 0.2130 0.2561 1.373 74 95.3

U-63 0.2128 0.2566 1.374 73 95.1 Stainless Steel Dummy Pellet 0.2197 in. 0.d. X 0.253 in. Long." .

Stainless steel dummy pellets acted as guides to center Tungsten-Rhenium thermocouples.

. . TABLE 8-3 FUEL COLUMN LENGTH

Measured Tabulated ' Length Length*

(Inches). (Inches)

Blanket . 1.025 - 1.0219 Fuel 5.978 5.9947 Insulator 1.022 1.0209

Total Length 8.026 8.0375

. Sum of individual pellet lengths from Table 8-2. Values ' . for the blanket and insulator pellets include the stainless steel

dummy pellets.

APPENDIX C

THERMAL CONDUCTIVITY CURVES FOR 304 SS, 316 SS, 321 SS, Zircaloy-2 AND NaK

Evaluation of data obtained from capsule ElAA re- was made to obtain values for these materials, and quired a knowledge of the thermal conductivity of 304 SS, polynomial expressions were derived which best fit the

316 SS, ,321 SS, Zircaloy-2, and NaK. A literature review available data. The polynomials used were as follows:

Applicable Temperature

Material Range (OF)

304 SS 150 < T < 1300 316 SS O

GEAP- 13603

APPENDIX E

DETERMINATION OF FUEL CLAD SURFACE TEMPERATURE

The fuel clad surface temperature was determined by

two methods. The first method involved using the thermal

conductivity of NaK. It was assumed that the temperature

recorded by a thermocouple next to the cladding was the

same as the temperature of NaK located at the geometric

center of that thermocouple. The second method involves

linear extrapolation of thermocouple temperatures to the

clad surface.

I t had prcviously been observed that the temperature

measured by thermocouples around the fuel cladding were

directly proportional to thc radius of the ther rnoc~u~ le .~

Assuming uniform radial heat dissapation from the fuel pin,

the data from 20 mil and 1116-inch diameter thermo-

couples at different angular locations next to the fuel

cladding showed as much as 50 to 60°F difference in

readings taken at the same time. Before data from the

20 mil and 1116-inch thermocouples could be compared

for a linear extrapolation of cladding temperature, this

angular temperature variation had to be accounted for.

A linear relationship was found to exist between a

fuel cladding thermocouple and the other thermocouples in

that set. That is, for any temperature measured by a clad-

ding thermocouple, there was always a related temperature

for the middle thermocouple and the outer thermocouple

in that set. In both sets of thermocouples employing a

1116-inch thermocouple next to the cladding, the same re-

lation was found to exist.* The linear relationship for the

case of a 1116-inch thet-rriocouple next to the cladding is

shown in Figure E-1 and for a 20 mil thermocouple in

Figure E-2. I t becomes evident that with the above graphs

the remperatures ot both a 20 mil and 1116-inch cladding thermocouple may be determined at any angular position in

the capsule where there wcre thermocouples. This is true regardless of whether or not a 20 mil or 1116 inch thermo-

couple actually existed at that angular position.

Using the above technique, a plot of temperature

versus thermocouple radius could be made, Figure E-3.

Extrapolation of the thermocouple data to a radius of zero

indicated the outside surface temperature of the fuel clad.-

ding. Agreement was excellent between cladding surface

temperatures determined graphically and temperatures

rigorously calculated using the thermal conductivity of NaK

across the thermocouple radius. Table E-1 shows tem-

peratures determined by both methods for 10 observations

in which the disagreement averages less than 2 3OF for a ,

temperature range of 720 to 1220°F. The significance of

this evaluation is the error involved if the cladding surface

temperature is assumed to be that of the thermocouples in

intimate contact with the cladding, e.g. -125'~ for

1116-inch diameter thermocouple at 1 1 0 0 ~ ~ . The error

appears to increase with temperature. For the purpose of

this report, the geometric technique was used to determine

fuel cladding surface temperatures.

As previously noted, there was an apparent cir-

cumferential variation in temperature in the capsule.

Thermocouple outputs showed that the four sets of

thermocouples followed a consistent pattern, i.e., one set of

thermocouples always read higher than another. The small

initial cold gap assured that fuel-cladding eccentricity would not be a factor. The variation is presumed to be

caused by the difference in angular orientation of the

thermocouples with respect to the reactor core. Considering

the near symmetry of cladding temperatures at positions

adjacent t o thermocouples 6 and 7, a hypothetical orientation of the capsule with respect to the core could be

determined, Figure E-4. Also included in the figure is a

table showing calculated cladding surface temperatures at

various times. At the higher temperatures the variation in

surface temperature recorded was as much as 50 to 60°F. I f

this i s extrapolated from the peak temperature, near

The capsule was regularly adjusted in the V-raft facility to assure thermocouple 18, to a point on the opposite side of the

that the thermocouples in the two horizontal planes were at the fuel cladding, a temperature gradient of as much as 10o°F

same power. may be postulated to exist around the cladding.

TEMPERATURE OF CLADDING THERMOCOUPLES (OF) (NO. 13 AND 18)

. f'

Figure E-1. Relation of Outer Thermocouples to 1/16 Inch Thermocouples Next to.Fuel Cladding

GEAP 13603

TEMPERATURE OF CLADDING THERMOCOUPLE (OF) . (NO. 6 ) '

Figure E-2. Relation of Outer Thermocouples to a 20 Mil Thermocouple Next to the Fuel Cladding

1

GEAP 13603

10 20 30

DISTANCE FROM CLAD WALL (Mils)

Figure E-3. Cladding Surface Temperature Determined Using a 20 Mil and 62 Mil Thermocouple

GEAP 13603

FUEL Dl AMETE R 0.22 in.

'DETERMINED FROM OUTER THERMOCOUPLES

THERMOCOUPLE DIAMETERS: NO.'$ 6 AND 7 0.020 in. NO.'s 13 AND 18 0.062 in.

-

tDATE AND TIME OF DAY , f. ttDETERMlNED BY LINEAR EXTRAPOLATION

Figure E-4. Cladding Temperature Related to Thermocouple Temperature

CENTER TEMPERATURE

1 1 88Oc 1 188Oc lltlllO~ 1 l8euC

1 BOB% 1660% l660Oc 1660'~

18730~ 1873'~ 1873'~ 1873'~

21 10°C 21 1 OUC

T.C. NO.

6 18 7

13

6 18 7

13

6 18 7

13

6 18

DATOD t

6-8, 24:W 6-8, 24:OO 6-4, 24:OO 6-8. 24:OO

6-1, 2u:w 6-7, 20:OO 6-7, 2O:OO 6-7, 2O:OO

6-10,15:08 6 1 0 , 15:- 6--10, 15:OO 6-10, 15:08

6-14, 4:QO 6--14, 4:OO

T.C. TEMPERATURE

658OF 639 665' 595 --- 987 945

1010' 900

1122 1084 1 150. 1015

1172 1110

CLAD TEMPERATUREtt

680'~ 701 686

- 654

1021 1043 1046 997

1162 1205 1191 1140

1213 1233

TC No.

TABLE E-1 COMPARISON OF FUEL CLAD TEMPERATURES

PREDICTED BY CALCULATION AND A GEOMETRIC APPROACH

Clad Temperature Clad Temperature by By Calculation (OF) Geometric Method (OF) Difference (OF)

FLUX.DEPRESSlON .FOR EIAA

In a thermal reactor, mixed oxide fuel exhibits the Isotopic Composition of Fuel:

phenomenon of self-shielding. The center portion of the ll-3.75 fuel i s partially shielded from the neutron flux by the outer

U-238 periphery. For this reason the local power generated at the center of the fuel pin i s lower than the power generated in the outer portion of the fuel. This in turn affects the cal- culation of integral conductivity of the fuel. To account for Pu-239 . , self-shielding a power shape factor (PSF ) was determined- Pu-240 for ElAA. The power shape factor i s a function of the neutron absorption cross sections of the mixed oxide and

Pu-241 ,

the geometry of the fuel pin. Pu-242

The neutron absorption cross section of the mixed oxide fuel was calculated as follows:

Absorption Macroscopic Cross Section (Za) for n isotopes:

Fuel Parameters : n

% Pu 20.28 ~a = z ~~o~~ ,-. : 92.3

i = 1 Density, % T.D.

U-235 9.9 Enrichment, %

U-238+U-235 where /I! Weight 33.48 gm N, = atom den'sity (atoms/cc)

I 2. Length 5.978 in. uai = microscopic cross section of one

Diameter

Annulus

0.2190 in.

0.073 in.

atom of isotope i

Ni = pi Na Wi

Ai

, where pi = density (gmlcc)

Na = 6.023 ( 1 0 ) ~ atomslmole Ai = atomic mass vi = volume fraction

pElAA = theoretical density mixed oxide Xfraction of theoretical density ot ElAA

pElAA = ( 1 1.0495 gmlcc) (0.9230) = 10.1987

PI i = (Fraction of volume is I ) (fraction of I is i)

\F~u-239 = (0.2028) (0.9068) = 0.183899

Vi\pu-240 = (0.2028) (0.0849) = 0.01 721 77

\Fpu.241 = (0.2028) (0.0079) = 0.001 6021 2

Vpu-242 = (0.2028) (0.0004) = 0.000081 12 vL'-238 = , (0.7972) (0.901 ) = 0.718277

\FU-235 = (0.7972) (0.099) = . 0.0789228

N i oai (cm)

Xa239 = 4.16839 ( 1 0 ) ~ ' (960110~~) 4.0416

Xa240 = 3.88834 (1 o ) ~ (254110~ ) = 0.09964 Ca241 = 3.60488 (10)' (1 165110~~) = 0.04236

Ca242 = 1.81859 (10)' ( 2 3 1 1 0 ~ ~ ) = 0.00004165

Ca238 = 1 .63413(10)~~ (2.3110~~3 = 0.038129

h235 = 1.81572 ( 1 0 ) ~ ' (572110~~) = 1.0485

Total Absorption = 5

[ = Ca r = (5.2188 cm-' (0.219012 in.) (2.54 cmlin.) = 1.4515

By using C on a Horwitz curve23 the self-shielding For this ratio a curve was determined which related the factor may be determined for various radial positions ratio of center annulus to fuel radius (0) to the power shape

factor (Figure ~ - 2 1 . ~ ~ . (Figure F-1). The ratio of the factors at fuel center and fuel surface

give the flux ratio of center to surface. d annulus - 0.333 = dfuel 0.2 190

PSF = 0.522 (From Figure F-2) -

GEAP 13603

Figure F-2. Power Shape Factor for E lAA

GEAP 13603

Figure F-1. E lAA Self Shielding Factor R = 0.1095 Inch

G EAP- 1 3603

ACKNOWLEDGMENT

The authors wish to acknowledge the assistance of

the following personnel in this experimental endeavor:

N. C. Shirley-experimental design, H. S. Currie, G. R. Irwin, and W. L. Hall-capsule fabrication; R. M. lrvine and R. E. Smith-post-irradiation examination; W. L. Conant for his help during the capsule irradiation and the post-

irradiation thermocouple calibration test. Finally, we would

like to extend our thanks to W. E. Baily, C. N. Spalaris, and E. L. Zebroski for their direction and recommendations.

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