Diffusion of fission product tritium em irradiated UO2

Journal of Nuclear Materials 74 (1978) 62-67 @North-Holland Publishing Company

DIFFUSION

OF FISSION PRODUCT TRITIUM IN IRRADIATED

U02

D. SCARGILL Chemical Technology Division,Atomic Energy Research Establishment, Harwell, Oxfordshire, U.K. Received 11 November 1977

The effect of temperature on the rate of release of tritium from neutron-irradiated polycrystalline and single-crystal UOa has been studied. The diffusion coefficients for tritium from SOO-1OOO’C are found to be several orders of magnitude lower than those reported for molecular hydrogen in UOz single crystals. Fractional release curves are indicative of classical diffusion and the Arrhenius plots for both polycrystalline and single crystal materials can be fitted by the diffusion equation: D = 0.12 (+0.21 - 0.07) X exp(-43 600 + 1800 cal/RT) cm*/,. L’effet de la temperature sur la vitesse de degagement du tritium provenant d’echantillons monocristallins et polycristallins de UO2 irradies par des neutrons a et6 etudid. Les coefficients de diffusion du tritium entre 500 et 1000°C ons eti trouves 6tre de plusieurs ordres de grandeur inferieurs a ceux publies pour l’hydrogene moleculaire dans des monocristaux de UO2. Les courbes de degagement sont typiques dune diffusion classique et des coefficients de diffusion i la fois pour des dchantillons mono et polycristallins obiissent a une loi d’Arrhenius d’equation suivante: D = 0,12 (+0,21 -0.07) exp(-43 600 f 1800 cal/RT) cm2/s. Der Temperatureinfluss auf die Freisetzungsrate von Tritium aus neutronenbestrahltem poly- und einkristallinen UOz wurde untersucht. Der Diffusionskoeftizient des Tritiums zwischen 500 and 1OOO’Cist einige Grossenordnungen niedriger als der Wert fur molekularen Wasserstoff in U02-Einkristallen. Die relativen Freisetzungskurven deuten auf eine klassische Diffusion hin. Die Arrhenius-Darstellung ftir poly- und einkristallines Material kann durch die Diffusionsgleichung D = 0,12 (+0,21 - 0,07) exp(-q/RT) cm2/s mit q = 182 + 8 kJ/mol dargestellt werden.

1. Introduction

poor penetration of tritium into the pellets, and the tritium was easily desorbed in a sweep of Ar4% H2 at 500°C. Uniform dispersion of tritium, however, was obtained using (U, Th)02 and (U,Ce)02 pellets to provide lattice vacancies for tritium; only adsorbed surface tritium was desorbed by a helium purge at 500°C for four hours. The behaviour of tritium in these mixed oxides is consistent with that in irradiated oxide reactor fuels which have been shown [4] to retain a greater fraction of the tritium generated by fission than would be expected from the hydrogen diffusion data. Moreover, heat-treatment at 55O’C in an inert atmosphere or under vacuum has been shown [5] to give only partial (3 l-58%) removal of the residual tritium from irradiated (UgU)O, fuel. This paper describes measurements made on

Tritium (t l/2 = 12.3 y) is generated in uranium oxide (UOs) fuel by ternary fission with thermal and fast fission yields [l] of about 0.9 X lOa and 2 X lo4 atoms per fission respectively for 235U. Previous studies on the diffusion of molecular hydrogen labelled with tritium tracer in unirradiated single crystals of UOz [2 1, in which the O/U ratio was 2000, indicate that the diffusion of tritium out of UOa should be rapid at elevated temperatures. These single crystals were impregnated by heating for several hours at SOO-7OO’C in an atmosphere of hydrogen plus tritium. Similar attempts to prepare tritium-impregnated UOz pellets for simulation of irradiated reactor fuel [3], by heating for five hours at 600-7OO’C in an Ar4% H2-3H2 atmosphere, gave 62

D. Scargill/ Diffision of tritiumin irradiatedUO2

lightly irradiated UOz to determine the effect of temperature on the rate of release of tritium produced in the UOz by fission. The work was carried out in order to (i) determine the conditions required to release >99% of the tritium from irradiated UOZ by heat treatment as a possible headend process for spent reactor fuel prior to dissolution at a reprocessing plant, and (ii) provide basic information for use in the computation of the fraction of tritium released from oxide fuels during reactor operation.

2. Experimental 2.1. Materials The polycrystalline, sintered UOZ (O/U < 2.003) and single crystal UOZ (O/U = 2.000) materials were irradiated in the Harwell DID0 Reactor at ambient temperature (50-9O’C) in a thermal neutron flux of about 6 X 10” n/cm2. Fragments from the same batch of polycrystalline material were irradiated for either two days (samples A) or fifteen days (samples B) in silica ampoules to burn-ups of 2.73 X 10” and 2.33 X 10” fissions/cm3 (about 0.001 and 0.01 at% burn-up respectively). Burn-up was determined by measurement of the total yield of “Kr generated in the sample using a fission yield for “KI [6] of 2.88 X 10m3 atoms/fission. Fragments from the same batch of single crystal material were irradiated similarly to a burn-up of 9.6 X 1017 fissions/cm3. The irradiated specimens were cooled for >200 days before use to allow for decay of 1311to an insignificant level. High-purity cylinder nitrogen was used in the preparation of a gas mixture also containing about 1.2 ~01%H2and 0.1 ~01%Kr. Tritium and 85Kr sources used for testing the trapping system were supplied by the Radiochemical Centre, Amersham. These gases were diluted with the abovementioned nitrogen gas mixture when in use. A standard “Kr calibration source was supplied by the National Physical Laboratory, Teddington. A known fraction of this source was adsorbed on a charcoal trap to give a suitable standard source (10.4 ctCi) for gamma counting with a NaI(Tl) scintillation detector.

63

2.2. Treatment conditions Weighed samples (20-100 mg) of irradiated UO2 were heated for 3 to 6 hours in a vertical quartz reaction tube (1 cm internal diameter) in the presence of a sweep gas introduced via a 2 mm bore quartz tube. Additional gas and the fuel specimens were introduced via a side-arm at the top (cool exit end) of the reaction tube. The treatment temperature was controlled to an accuracy of +2”C and measured by a precalibrated thermocouple strapped externally to the reaction tube. Water vapour (about 2.2 ~01%)was introduced into the gas streams when required by bubbling them through water at room temperature. Either nitrogen (with or without water vapour) or the gas mixture (1.2 ~01%H2 and 0.1 ~01%Kr in nitrogen) was used as the sweep gas; in the former case, the gas mixture was added to the exit gas so that hydrogen could act as a carrier for any elementary tritium; in the latter case, water vapour was added to the sweep gas at the end of the experiment to ensure transport of any adsorbed tritium oxide to the offgas cold trap. Gas flows of lo-70 cm3/min were controlled by a rotameter-needle valve system. Stainless steel tubing (1.5 mm i.d.) was used to conduct the inlet gases to the reaction tube, and the outlet gases via a millipore filter to the gas trapping system. Gas inlet pressures were kept at about 0.7 kg/cm2 and the pressure in the system was 0.1-0.2 kg/cm2 depending on the flow rate. 2.3. Trapping and analytical procedures The order of removal of the volatile fission products from the off-gas stream was tritium (as HTO), tritium (as HT) and krypton-85. Tritiated water vapour (HTO) was separated from the off-gas stream by freezing out in a glass trap (1.5 cm i.d. X 18 cm long) cooled to -78°C with a dry ice-acetone slush. Tritiated hydrogen (HT) was oxidised to HTO by passing through a bed of copper oxide (about 7 g) heated to 630°C (complete oxidation was found to occur at about 560°C in separate tests) before it was trapped out in a second cold trap. These traps were renewed at noted time intervals during the experiment. Each trap contained about 0.5 cm3 of added water before use in order to mini-

64

D. Scargill /Diffusion

mize losses of HTO on warming the frozen cold traps to room temperatu~. After removal of this liquid, the traps were rinsed out twice with 1-2 cm3 water and the combined liquids analyzed for tritium by liquid scintillation counting of 1 cm3 aliquots. Absolute disintegration rates were obtained by comparison with a standard solution of tritiated water. Residual tritium (and 8sKr) in the treated fuel was determined by total dissolution under reflux conditions in l-2 cm3 of 6 M HNOs at 14O“C.For determination of any tritium evolved as HT, the offgases were swept out of the reflux dissolver with a stream of the gas mixture containing hydrogen, bubbled through about 20 cm3 of 5 M NaOH solution (to remove acid vapours and radioactive spray) and passed via a cold trap (to remove water vapour) to the heated copper oxide-cold trap train. In order to analyse for tritium in the dissoIver solution pius condenser washings, it was found necessary to carry out a triple distillation procedure to eliminate fission product contamination. Before distillation, the initial solution containing uranium was neutralised, and the intermediate condensates were made alkaline (about 0.1 M in NaOH). Tests on simulated solutions containing a known amount of tritium, established that this procedure gave yields correct to *3% when 6040% of the liquid was distilled at each stage. “Kr was removed from final dissolution off-gas streams by passing the gas through a series of tubular cold traps (-‘78’C) each containing about 12 g of charcoal. For mounting purposes, the adsorbed krypton was transferred to a flat cylindrical trap containing about 7 g of charcoal. The 85Kr released was used to monitor the progress of dissolution by means of an in-line gas flow scintillation counter, and the total yield (Table 1) was used to calculate the fission yield of the tritium (found about 8 X 10e4 atoms/~ssion). Less than 0.15% of the total *‘Kr found was released on heat-treatment for 5 hours at 810°C.

of tritium in irradiated lJO2

with a flow rate of 66 cm3/min and about 1 ~01% Ha0 vapour, the time required for tritium activity to reach an in-line gas flow scintillation counter via about 1 m of stainless steel tubing was <2 min, but 10 min was required to reach the true activity of the initial gas. With a dry glass tritium trap at room temperature inserted in the line, the times were 14 min and 70-80 min respectively. This adsorption behaviour was reflected in relatively large and variable activities found initially in pre-experiment blanks for the off-gas tritium traps, even after purging the apparatus with dry argon for 24 hours. Purging with nitrogen gas saturated with water vapour gave appreciably lower blanks. Tritium as HT (and also “Kr) passed through the off-gas system in about 2-4 min at the flow rates used in the experiments. 3.2. Form of release of t~tium on heat ~eatme~t atrd dissolution The form of tritium (HT or HTO) released on heating was found to be dependent mainly on the form of hydrogen (Ha or HsO) in the sweep gas, indicating isotopic exchange (see table 1). Thus with water vapour in the sweep gas, only relatively minor proportions of tritium as HT were found in the offgases, whereas with pure nitrogen or nitrogen containing hydrogen as the sweep gas, only 20-26% of the tritium released was as HTO. The presence of HTO in the latter experiments can probably be ascribed to the presence of traces of water vapour in the cylinder gas and/or pick-up of moisture from the gas inlet lines. Dissolution of UOa released a negligible fraction (0.1 to 1%) of tritium as HT in confirmation of other results [4,7] for oxide fuels. The amount found as HTO, where measured in the caustic soda off-gas trap, was equal to that expected by carry over of water vapour in the saturated sweep gas.

3. Results

3.3. Rate of refease of tn.tium

3.1. Transport behaviour of tritium in the apparatus

The rate of release of tritium during heating of the UOs was determined by measurements on the principal form evolved (either HT or HTO). It was assumed that the two forms were emitted in constant proportions in each experiment. Tritium yields

The rate of transport of tritium as HTO through the apparatus was found to be dependent upon the water vapour con~ntration in the gas stream. Thus,

65

D. Scargill /Diffusion of tritium in irradiated UO2 Table 1 Release of tritium Sample

from irradiated

UOz by heat-treatment Tritium

treatment

released

UOz (sample

525” in 1.2% Hz-N2 (A) 625” in 2.2% H20-N2 (B) 675” in 1.2% Hz-N2 (B) 810” in N2 (B) 810” in 2.2% H20-N2 (B) 1000” in 2.2% HzO-N2 (B) Dissolution (A) Dissolution Single-crystal

UO2 (Average

oJZi/g UOz)

85Kr

<2

0.08 1.01 0.66 0.86 0.91 1.02 0.103 0.87

ND ND ND 36.5 30.7 ND 3.9 ND

0.63 f 0.05 2.01 + 0.1 7.55 + 0.2 ND

0.353 ND 0.40 0.36

13.3 ND ND 14.4

A = 0.1 PCi T/g, sample B * 0.9 nCi T/g) 0.47 0.08 5.1 11.2 0.93 13
+ * + f f f

0.2 0.02 0.2 0.2 0.2 0.3

0.21 3.6 1.5 2.9 18.9 84
<0.5

(B)

found

T

HTO

HT Polycrystalline

Total activity

(% of total found)

f 0.1 f 0.1 ?l k 0.3 f 0.5 *2

= 0.37 PCi T/g)

650” in 2.2% H20-N2 750” in 2.2% HzO-NZ 850” in 2.2% H20-N2 Dissolution

0.1 r 0.05 0.06 f 0.03 0.7 f 0.1 0.6

ND = not determined.

(/.&i/g) averaged for each batch of material were used for normalisation of the fraction of tritium released as a function of (time)r~. This gave the typical diffusioncontrolled release plots shown in fig. 1 for the polycrystalline material and in fig. 2 for the single crystal material. The straight line relationships shown for the polycrystalline material were obtained after correction of initial plots so as to show zero release at zero time, thus allowing for the dead time of the apparatus. These results indicate that the rate of release is independent of the carrier gas used (Hz or H20) in the sweep gas. Only at 1000°C was the fraction of tritium released sufficiently large to show a marked deviation from the linear relationship. The rate of release curves shown in fig. 2 for the single crystal material indicate that there was a small burst of release at the start of the experiment.

o..N2+2 0 18

A N2.12 o

016

2 vol%

H20

vol%H2

N2 as sweep

OS sweep (1s sweep

90s o-9

gas

gas

0 IL

07

: VI 2 012 i CL

06

$ ;

010

05

008

OL

: z

3.4. Diffusion coefficients of tritium The diffusion equation [8] relating the cumulative fraction f of gas atoms released in time t to an infinite void with the diffusion coefficient D is given by f = 2 (Dt/rr)rP

S/V,

(1)

2

‘

6

a

10

i TIME I MIN

Fig. 1. Release of tritium

II '/z

12

from irradiated

11

16

polycrystalline

18

UO2.

D. Scargill /Diffusion of tritium in irradiated UO2

66

TIOCI

1

0 06

10-6,

1000 900

800

700

600

500

007.

2

L

6

6

10

ITIMELMINII

,, 12

IL

16

16

2

Fig. 2. Release of tritium from single-crystal UO2. L

7

8

9

10

11

12

13

1L

10L/TloK-'1

where S/V is the surface-to-volume ratio of the specimen. For small values off, this expression may be related to the diffusion from a finite body which is often assumed hypothetically to be a sphere of radius a, using the relationship S/V = 3/a.

D = 0.12 (tO.21-0.07) cm2/s.

(3)

flt ‘I2 is then obtained

from the slope of the fractional release vs. t’p curves (figs. 1 and 2). The value of a for the polycrystalline material was calculated from the measured BET surface area (about 10 cm’/ g) to be about 0.03 cm. The value of S/V for the single crystal material was calculated by assuming cubic geometry for the fragments. Using the above parameters for the two types of material, values of D were obtained as a function of temperature T(K) and gave a straight line relationship when plotted as an Arrhenius diagram of log D vs. l/T (fig. 3). From the Arrhenius relationship D = Do exp(-Q/RT),

the values of Do and the activation computed to give the expression

energy Q were

(2)

Eq. (1) is valid for single particles or a number of identical particles and for a wide variety of particle shapes, provided the f< 0.1. It can be rewritten as f/t ‘I2 = 6 (D/na2)‘j2,

Fig. 3. Diffusion coefficients of tritium in irradiated UO2.

(4)

exp(-43600

f 1800 cal/RT) (9

4. Discussion 4.1. Diffusion behaviour of tritium in irradiated U02 Diffusion coefficients for tritium are compared with those reported [3] for hydrogen in fig. 3. Diffusion coefficient values obtained in this work for fission-product tritium in UO2 are lower by factors of 4 X 10’ to 2 X lo4 over the temperature range 500 to 1000°C respectively than those found for molecular hydrogen, and the activation energy for tritium diffusion (43.6 kcal/mol) is much greater than found [3] for hydrogen (14.3 kcal/mol) presumably because of the different species involved (either atomic tritium or molecular hydrogen form).

D. Scargill /Diffusion of tritium in irradiated UO2

The insensitivity of the rate of release of tritium to the presence of hydrogen or water vapour in the sweep gas suggests that these substances are unable to diffuse to the lattice sites of the tritium, since isotopic exchange should be rapid at the temperature used. Exchange will take place, however, when the tritium reaches a crystal surface. The precise mechanism for the diffusion behaviour cannot be given, although one may postulate formation of e.g. -OH groups in association with some of the fission-product impurties, the atoms of which are in far greater abundance than the tritium atoms.

61

(3) The activation energy for fission-product tritium diffusion in UOz (43.6 kcal/mol) is much greater than for the molecular hydrogen. (4) Diffusion release of tritium is the same from single crystals and sintered polycrystals of UOa. (5) Heat-treatment at about 700°C could be a feasible technique for the release of tritium from spent reactor oxide fuels, provided that (a) the particle radius of the restructured fuel is not greater than about 2 X low3 cm, and (b) the diffusion behaviour of tritium is not changed significantly at high burn-up of the fuel.

4.2. Removal of tritium from spent reactor fuel Acknowledgements Studies at Oak Ridge (USA) [S] have shown that oxidation of UOz in oxide fuels to UsOa by heating in air at SOO-600°C gives rapid and nearly complete release of tritium in a few hours. This treatment is being actively considered for the release of residual tritium from oxide fuel prior to dissolution. The time required to give rapid release of tritium by heating the fuel, without oxidation, will be dependent on the particle size of the fuel which is related to the degree of restructuring during irradiation. For desorption of gases from a sphere [9] the time t for >99% release is related to the diffusion coefficient D and the particle radius a by the equation Dtfa’ = 0.5. Assuming a particle radius not greater than 2 X 10e3 cm for the spent reactor fuel, calculations indicate that a temperature of >7OO”C would give >99% release of tritium in
5. Conclusions (1) Fission-product tritium is released as hydrogen gas from UO2 in the absence of water vapour. (2) The diffusion coefficients determined for fissionproduct tritium in UOa are lower by several orders of magnitude than those reported found for hydrogen in U02.

The author thanks Dr. F.A. Johnson and J.R. Findlay for helpful discussions on interpretation of diffusion data, Dr. H.A.C. McKay for his support throughout the course of this work, M.J. MoretonSmith for supplying the uranium dioxide materials and providing information on the BET Surface area measurements, and T.J.H. Elliott for experimental assistance.

References [l] M.J. Fluss, N.D. Dudey and R.L. Malewicki, Phys. Rev. C6 (1972) 2252. [2] V.J. Wheeler, J. Nucl. Mater. 40 (1971) 189. [3] J.H. Goode, C.L. Fitzgerald, G.D. Davis and O.L. Kirkland, USAEC Report ORNL-TM-4141 (1972) 13. [4] J.H. Goode and V.C.A. Vaughan, USAEC Report ORNLTM-2793 (1970). [5] J.H. Goode and V.C.A. Vaughan, USAEC Report ORNLTM-3723 (1973) 115. [6] F.L. Lisman, R.M. Abernathey, R.E. Foster and W.J. Maeck, J. Inorg. Nucl. Chem. 33 (1971) 643. [7] J.H. Goode, USAEC Report ORNL-3956 (1966). [ 81 W. Inthoff and K.E. Zimen, Trans. Chalmers Univ. Technol., Gothenburg No. 176 (1956). [9] The Mathematics of Diffusion 2nd Edition J. Cranked., (Clarendon Press, Oxford, 1975).