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The chemical effects of composition changes in irradiated oxide fuel materials II-Fission product segregation and chemical equilibria
Journal of Nuclear Materials 61 (1976) 0 North-Holland Publishing Company
254-270
THE CHEMICAL EFFECTS OF COMPOSITION CHANGES IN IRRADIATED OXIDE FUEL MATERIALS II-FISSION PRODUCT SEGREGATION AND CHEMICAL EQUILIBRIA F.T. EWART and R.G. TAYLOR UKAEA, Applied Chemistry Division, Atomic Energy Research Establishment. HarweU, Oxfordshire OX1 1 ORA, lJK
J.M. HORSPOOL and G. JAMES UKAEA, Research and Development Laboratories, Windscale Works, Sellafield, Cumbria CA20 IPG, UK Received
9 February
1976
Four miniature fuel pins containing uranium/plutonium oxide, covering a wide range of oxygen-to-metal ratios, have been irradiated in a thermal reactor. These pins were subjected to a detailed examination, by electron probe microanalysis and ceramography. The results of this analysis have been interpreted in the light of the results of a computer model of the pin behaviour . The widely different fuel chemistry in each of the fuel pins produced different fission-product phases and different clad corrosion phenomena. The phases found are in general agreement with the model which predicts increasing oxygen-to-metal ratio with increasing burn-up and oxygen redistribution in the temperature gradient. Clad corrosion is seen to be inhibited at low oxygen potentials but the distinction between intergranular and ‘broad front’ corrosion is not attributable
simply
to the local oxygen
potential.
Quatrc epingles de combustible miniatures contenant de l’oxyde mixte d’uranium et dc plutonium, couvrant un large domaine de rapport oxygene/metal, ont Cte irradiees dans un reacteur thermique. Ces dpingles furent soumises i un examen detail16 par microsonde Clectronique et ceramographie. Les resultats de cette analyse ont dtd interprdtes a la lumierc dcs rcsultats d’un modele de calcul sur lc comportement des aiguilles de combustible. La chimic du combustible tres differentc pour chacune des aiguilles combustibles avait pour consequence de produirc diffdrcntcs phases de produits de fission et divers phenomenes de corrosion du gainage. Les phases trouvecs sont en general d’accord avcc le modtle qui prdvoit un accroissement du rapport oxygene/metal avec le taux de combustion et la redistribution de I’oxygenc dans le gradient de temperatures. La corrosion du gainage apparait &tre inhibee aux faibles potent& d’oxygene, mais la distinction entre corrosion intergranulaire et ‘front general’ n’est pas attribuable simplement au potentiel total en oxygene. Vier kurze Uran-Plutonium-Oxid-Brennstabc mit schr unterschiedlichem Sauerstoff/Metall-Verhahnis uurden in cinem thcrmischcn Reaktor bestrahlt. Die Brennstabe wurden keramographisch und mit der Elektronenstrahlmikrosondc eingehend untersucht. Die Ergebnisse diescr Analyse wurden unter dcm Aspckt der Ergebnisse eincs Rcchenmodells fir das Brennstabvcrhalten gedeutet. Die sehr unterschiedliche Chemie des Brennstoffs in jcdem der Brennstabe hat unterschiedlithe Spaltproduktphasen und Htillkorrosionsphanomene zur Folge. Die beobachteten Phasen stehen in ungefahrer Ubereinstimmung mit dem Modell, das tin steigendes Sauerstoff/Metall-Vcrhaltnis mit dem Abbrand und tine Sauerstoff-Umverteilung im Temperaturgradicntcn voraussagt. Die Hiillkorrosion ist bei niedrigen Sauerstoff-Potentialen gehemmt, der Unterschied zwischcn Korngrenzenund flachcnhafter Korrosion kann nicht einfach auf das lokale Sauerstoff-Potential zurtickgeftihrt werden.
fuel by a range of fission product species were discussed and the likely chemical form of the principal fission products was considered. Supporting experimental evidence was presented and it was concluded that there are two principal effects: firstly, that the average valency of the fission products is less than that
1. Introduction In an earlier paper [ 11, experimental evidence was presented which led to conclusions about the chemical state of an irradiated oxide fuel. Chemical modifications caused by the replacement of fissile atoms in the 254
I? T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
of the fissile species such that the fuel becomes more oxidising with burn-up; and secondly, that some of the fission products will form separate phases which affect the overall oxygen balance and may modify the performance of the oxide as a reactor fuel. An additional effect that was not considered in these earlier experiments, which were under near isothermal conditions, was the extremely steep temperature gradient which obtains in an operating fast reactor type fuel pin (about 800 K/mm). The principal effect of this gradient in the oxide fuel system is to provide the driving force for oxygen redistribution. This, in turn, has an important bearing on the operational performance of a fast reactor fuel pin and the optimum choice of oxygen-to-metal ratio for the initial fuel. The magnitude of oxygen redistribution effects has been the subject of a number of theoretical assessments. A few experimental measurements have also been made; both of these aspects have been reviewed by Bober and Schumacher [2]. Experimental work on the phases formed in fuel during irradiation is described by Bramman and Powell [3]. Results are presented here from the irradiation of four miniature fuel pins which were conducted under carefully controlled conditions and then subjected to a detailed post-irradiation examination using electron probe analysis and other techniques. The initial oxygen-to-metal ratio of these pins was selected to cover a wide range, from O/M = 1.94 to O/M = 2.15. This provided an opportunity to relate the previous theoretical and experimental studies to actual fuel pin irradiations and to illustrate the effect of the thermochemical changes and the differences in fuel behaviour relating to the different conditions of oxygen-to-metal ratio. The limits of desirable oxygen-to-metal ratio for an irradiated fuel system are set in the upper range by compatibility with the stainless steel [4] clad and, in the lower range, by the formation of free metal in the fuel. Under isothermal conditions the range between these limits is large but in the operating fuel the oxygen redistribution in the temperature gradient and the change in oxygen-to-metal ratio during burn-up may make it impossible for the fuel not to exceed these limits during its operational life, depending on the magnitude and direction of the redistribution of oxygen and the rate of change of the oxygen-to-metal ratio.
255
The behaviour of the fuel pins under these extremes of initial fuel specification has been followed principally by the identification of fission product phases and other structural features observed in the fuel. These observations have been related to the local chemical conditions within the fuel, using a computer model describing the chemical state of the fuel. Whilst uncertainties exist both in definition of conditions and the mechanisms controlling the behaviour, it is clear that widely differing chemical conditions existed during the irradiations which were reflected in the phases developed within the fuel.
2. Experimental
procedure
The fuel within the four fuel pins for irradiation was derived from two batches of powder manufactured by a common route. The powders were then compacted and sintered using identical methods and then adjusted to three different oxygen-to-metal ratios. Each fuel pin contained approximately 12 grams of (U, 85Pu0 15) oxide to the following compositions: Pin 1 O/M = 1.94; carbon 1550 ppm Pin 2 O/M = 2.15; carbon
Details of the fuel are given in table 1, the irradiation conditions are in table 2. Each pin was 175 mm long, 5.84 mm outside diameter and 5.08 mm inside diameter. The clad was 20% cold-worked M3 16 stainless steel. The fuel stack was contained between two UO2 pellets and was retained by a collet, leaving a gas plenum of 70 mm. The irradiations were carried out in a static sodium-filled capsule for 16 reactor cycles in the PLUTO Materials Testing Reactor at Harwell. The nominal fuel rating was 200 W/g and the clad mid-wall temperature was 650°C. The samples of fuel were taken for electron probe micro-analysis (EPMA) from positions in the fuel pins shown in the cutting diagram in fig. 1. The transverse sections were mounted in conduct-
256
I;. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
Table 1 Characterisation
of fuel before
irradiation 99312
993/l Pu content Isotopic ratios
&
15.2
15.23
15.23
~238
0.73 99.27
0.03 0.89 99.08
0.03 0.89 99.08
Pll239 Pu2‘+0 PUZ4’
93.54 6.00 0.46
93.43 6.08 0.49
93.43 6.08 0.49
(
U
23h
) ’
-
u235
1555 < 100 170 400 260 80 < 100 660 < 100 < 10 -
Carbon cu Ca Al Mg Cr Mn Fe Ni Cd Zn O/M Sintering
99313
1.94 1 hr/1550 C/H2
conditions
Density % Theoretical
< 50 100 140
epoxy resin in a short bakelite cylinder and reduced in thickness to about 125 Mm thick, when the surface was polished to a 1 pm finish. Throughout these operations, nonpolar solvents were used to pre-
C/CO2
2.000 3 hr/1600 C/H2
99.2
serve the fission product caesium. At this thickness the radiation levels from the samples were sufficiently reduced to be acceptable to the microprobe analyser. The longitudinal samples were taken from the sam-
ing
Table 2 Details of irradiation
Two pellets of 993/2 in centre of stack of 993/3
50
98.2
95
99314
conditions Pin no.
Mean rating (w/g) Linear rating (kW/m) Irradiation time (MW/h) Average flux (n/cm’ . s) Can nominal midwall temp (“C) No. of reactor cycles Burn-up (%) Gas release (%)
1
2
3
4
165 29.0 196704 3.28 x 1Or3 650 16 7.3 63.7
183 33.5 196252 3.82 x 1013 650 16 8.1 71.5
199 36.4 200577 4.24 x 1013 650 16 9.2 82.1
172 31.8 196668 3.41 x 10’3 650 16 7.6 79.1
251
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materials11 993/L/01 9931212A
9931112A 993/3/2A
I
I
End
cop
Insulator
I
High O/M pellets in 9931.L
Approx
993/1/1A \993/2/1A
‘=/3/1A, I
I
Insulator
10 cm
Fig. 1. Diagrammatic section of fuel pin showing sections removed for analysis.
ples mounted for ceramography. These had been obtained by reducing a cylinder of fuel to a semi-cylinder. A slice 2 mm thick and 6 mm square was taken from the face of the ceramographic sample and was mounted and ground as previously. The bulk removal of material was from the rear of the sample so that the face finally observed was within the 125 pm of the original. The electron probe micro-analysis was carried out on a Cameca MS85 and a Cambridge Microscan 5 instrument. The development and shielding of these instruments has been described elsewhere [5]. The detailed searching of the sample surface for fission product segregation was carried out on the Cameca instrument which had a detection sensitivity under search conditions of 0.5 wt% of noble metals or clad constituents and 1 wt% of rare earths. Under the conditions used, the limit of detection when analysing on a fixed point was 0.1 wt%. It has been found that with these instruments it is not possible to obtain quantitative data from inclusions smaller than a 5 pm diameter hemisphere, because of errors introduced from matrix contributions due to the spread of the electron beam. However, relative concentrations may be estimated with reasonable accuracy for smaller inclusions. Where these are presented in the results, the ratios are those of the Xray intensities, which are approximately proportional to the weight fractions of the element. The quantitative data presented have been collected from uniform, dense, polished areas exceeding 5 pm in diameter. The data were reduced and corrected using the computer programme DMPEMP8T [6]. This programme presents the results as weight percentages and subsequently as atomic percentages normalised to 100%. Results were accepted for normalisation only when the range of the weight percentages was between
98% and 102%. Analyses falling outside this range were repeated under improved conditions or further elements were sought.
3. Results 3.1. Electron probe analyser observations 3.1.1. Pin 1 Two sections were examined from this pin, the transverse section 993/1/2A was taken from the centre of the fuel stack and the longitudinal section 993/ l/ 1A was taken to include the upper (plenum end) insulator pellet. The transverse section (fig. 2) contained two dif-
Fig. 2. Transverse section of pin 1.
X
12.5.
258
F.T. Ewart et al. f Composition changes in irradiated oxide fuel materials II
ferent metallic inclusions. The well known noble metal inclusion was found in the central void; this had the composition: MO 37.5, Tc 9.8, Ru 36.7, Rh 11.1; Pd 4.9 at%. Inclusions of similar qualitative analysis were found in the columnar grain region. A different type of metallic inclusion was found in a number of positions about 120 pm from the outer edge of the fuel (fig. 3). These were of a diffuse form and contained U, Pu and Pd in the approximate composition (U, Pu)Pd,. The composition was not, however, constant and some analyses obtained were: U 15.3,Pu
11.4,Pd73.3
at%
U 15.6, Pu 13.2, Pd 71.2 U 17.8, Pu
7.7, Pd 74.5
The fuel was examined for concentrations of clad species but norie was found. The fuel clad gap contained some well defined deposits which were found to contain only caesium. Caesium was also found in the fuel rim but was not found in the clad. The longitudinal section showed the presence of apparently metallic particles at the interface between the insulator pellet and the fuel stack. These inclusions contained U, Pu, Pd, Sn and the composition varied along the fuel radius. Approximate compositions (*5%
Fig. 3. PuPd3 type inclusion
pin 1. X 80.
on each component)
in atomic percentages
were:
Outer U1S,6Pd67.5Sn16.9 at% Inner Ug,3Pu4.1 Pdj8 3snS1 3 at% . 3.1.2. Pin 2 Two sections were examined from this pin, a transverse section 993/2/2A taken from position shown in fig. 1 and a longitudinal section 993/2/1A inciuding the plenum and insulator pellet. A micrograph of the longitudinal section is shown in fig. 4. The prominent features of both these sections were the extensive clad corrosion (fig. 5) and, in the longitudinal section (fig. 4), the complex group of phases found at the insulator-fuel interface. The observations will be presented in three groups, fuel matrix, insulator interface, and clad interface. Fuel matrix The surface microstructure features of the fuel matrix were similar to those commonly found in high burn-up oxide irradiations. White metallic inclusions
Pig. 4. Fuel-insulator x 12.5.
region,
longitudinal
section
pin 2.
F. T. Ewart et al, / Composition changes in irradiated oxide fuel mare&Is II
FULL Fig. 5. Clad corrosion
pin 2. x 280.
were found in the equiaxed and columnar grain region. A second type of metallic inclusion, darker than the first, was found in the fuel rim. The white inclusions were shown to be the expected noble metal inclusion but molybdenum was found to be absent; typical compositions for inclusions in the fuel centre position were: Tc 8.5, Ru 63.1, Rh 21.0, Pd 7.4 at% Tc 8.8, Ru 62.5, Rh 19.8, Pd 8.9 at% The Tc : Ru ratio in the inclusions was measured at several points across the fuel radius, and was found to be 1: 7 in the central void 1 : 11 in the columnar grains 1 : 7 in the equiaxed grains and 1 : 4 in the rim. The fission-product molybdenum was found only in the fuel matrix. The darker inclusions which occurred up to 200 pm from the clad contained U, Pd, Te in the approximate atomic ratio 1 : 6 : 3. Insulator interface
The area shown in fig. 4 is a longitudinal section of the interface between the fuel and the UO, upper in-
259
sulator pellet. The dominant features of this section are the hemispherical crack, up to 0.5 mm wide, which had formed in the U02 insulator pellet and the apparent extrusion of the insulator into the central void of the fuel stack. This section was examined by arautoradiography and EPMA to attempt to identify the causes of the hemispherical crack. No significant diffusion of plutonium into the insulator could be found and the penetration of fission products into the insulator was bounded by the crack. It is postulated, then, that the crack represents the limit of oxygen diffusion into the insulator or a temperature boundary. The material in the central void consisted partly of the extruded insulator and partly of the fuel forming a bridge. The section contained a large number of separate phases. The bridge contained a large inclusion, some 300 E.trnin diameter, which had the composition Pd 73.7, U 4.0, Te 22.3 at%. Smaller inclusions of similar composition were found in the fuel rim both in this section and in the transverse section of this pin. In this region, these inclusions were often associated with noble metal inclusions and a darker phase which :ontained only caesium, presumably as the oxide, as hewn in fig. 6. Where the extruded insulator pellet met the fuel bridge an iron and chromium-bearing phase containing only traces of nickel was found. A further phase was found in the insulator pellet about 1 mm below the hemispherical crack. The only metals found were MO and Tc at 10 and 57 wt% respectively, leading to the conclusion that the phase was non-metallic, the most likely form being a mixed oxide phase. About 100 pm from the crack and within 200 m of the clad a non-conducting, comparatively lowmelting phase was found. These properties were shown by the charging of the sample surface and the local melting which occurred when the electron beam was applied to this area. This phase, which contained Cs, Cr, Fe and MO, was thought to be a caesium chromatelmolybdate. In the same radial position as the previous phase, the hemispherical crack contained a complex structure of grain boundary phases which were too small for unequivocal analysis. It could be shown that the grain boundaries contained Fe, Cr and Ni concentrations, however, and were visually similar to the phase found at the clad interface.
260
F. T. Ewart et al. 1 Composition changes in irradiated oxide fuel materials II
which was rich in caesium. In some places on the section this phase was in the ‘gap’, in others it merged into the next layer. (c) A layer of corrosion product up to 75 pm wide. This layer contained Fe, Cr and Ni, the chromium fraction being greater, by a factor of about two, than in the clad itself. To a lesser extent, U, Pu, and Cs, Ba were also found. (d) A complex zone containing a number of bubbles and bounded on the inner side by a row of metallic inclusions. The zone consisted of alternate layers of zone (c) and of fuel. The metallic inclusions were difficult to analyse quantitatively, being less than 5 I.crnin diameter, but were found to contain Pd, Te and some Cs. (e) Substantially unaltered fuel which had in a few places been penetrated by phase (c) to a depth of 100 pm.
Fig. 6. Fission product inclusions pin 2 Pd-Te (light), Metals (bright) and Cs (dark). X 350.
Noble
Fuel-clad interface The clad corrosion on this pin was clearly of the ‘broad front’ type. The maximum clad penetration found was 75 pm; this was observed generally in the longitudinal section and more locally in the transverse section. The reaction product had a slightly different appearance in the two sections. The longitudinal section showed a reaction product attached to the clad and a reaction product on the fuel side of the gap. The transverse section, however (fig. S), showed the clad in contact with the reaction products, but in other places there was a gap between the clad and the reaction products. The examination of these features by the electron probe showed that in all cases the reaction products were the same. These are conveniently described sequentially from the clad to the fuel: (a) Undisturbed clad showing no variations in composition from outer to inner wall and no evidence of second phases in the grain boundaries. (b) A narrow, 10 ym, layer of variable density
3.1.3. Pin 3 Again both transverse 993/3/2A and longitudinal sections 993/3/1A were taken from this pin. The clad in both cases showed pronounced clad attack; this was quite different from the broad front attack seen in pin 2. Almost all regions where the clad was in contact with the fuel showed attack by an intergranular
Fig. 7. Clad corrosion
pin 3.
X
280.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
261
cant decrease in molybdenum content of the inclusions towards the centre of the pin. The values obtained were: Void edge, MO : Ru = 1 : 4
JNCLtiSlON
Fig. 8. Metallic inclusions in pellet interface pin 3. X 280. penetration mechanism (fig. 7). There was no evidence of the complex fuel-clad component phase found in the previous pin, although a metallic phase was observed in a few places at the pellet interfaces (fig. 8). The fuel was of normal appearance with metallic inclusions present in the columnar and the equiaxed grain regions. The elements present in the grain boundary phase in the clad were found to be caesium and the cladding constituents Fe, Cr and Ni. The caesium concentration was reasonably constant throughout this phase whereas the relative atomic concentration of the clad constituents varied widely, e.g. Fe 6-17, Cr 0.8-1.1, Ni 1. There was no variation in the concentrations of the clad constituents across sections of sound cladding. The apparently metallic phase found in the interpellet gaps (fig. 8) was shown to contain only Pd and Fe. In the sample presented it was not possible to carry out an absolute analysis, but the relative intensities corresponded to a composition between FePd and Fe,Pd. Inclusions containing clad components were found in no other locations in the fuel. The metallic inclusions found in the fuel were solely of the noble metal type. None of these inclusions was large enough for quantitative analysis. It was possible, however, to measure the relative concentration of molybdenum and ruthenium. There was a signifi-
Columnar,
MO : Ru = 1 : 3
Equi~ed,
MO : Ru = 1 : 1.8
Rim,
Mo:Ru=
1: 1.4
3.1.4. Pin 4 One transverse section 993/4/Dl for EPMA was taken from a position in the stack of originally stoichiometric pellets. The ceramography of a longitudinal section covering the region where pellets of differing oxygen to metal ratios were located suggested that equilibration had occurred between the two pellets, since the microstructures on each side of the pellet interfaces were identical. This was in contrast to the interface between the fuel and the insulator in pin 2. No evidence was seen on the section taken for EPMA of any clad attack; a slight intergranular penetration was found on other sections (fig. 9). The section contained a deposit in the fuel-clad gap. The fuel contained a number of metallic inclusions; in addition
Y
Fig. 9. Clad corrosion pin 4. x 350.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
262
to the bright noble metal type, a darker type of inclusion was found. Both types of inclusion were widespread in the fuel. As was expected from the visual examination, the electron probe did not detect any clad penetration by foreign elements and there was no evidence of composition variations across the clad. The deposit in the fuel-clad gap was of indefinite density and could not, therefore, be fully analysed; it contained predominantly caesium and barium with lower amounts of molybdenum and cerium, with traces of tellurium and neodymium. The two types of metallic inclusions found in the matrix were analysed by EPMA. The small bright inclusions found in the columnar and equiaxed grain region were the conventional noble metal type containing MO, Tc, Ru, Rh and Pd. As found in pin 3, these were too small for full analysis but the MO : Tc : Ru atomic ratio was measured as a function of radial position:
Void edge, MO : Tc : Ru = 1 : 1.4 : 6.0 Columnar,
MO : Tc : Ru = 1 : 1 .l : 4.2
Equiaxed,
MO : Tc : Ru = 1 : 1.7 : 9.3
Palladium was also found in the darker inclusions which were much larger than the noble metal inclusions and could be analysed quantitatively. Analyses of these inclusions taken at central and rim positions were: Centre, Pd 73.8, U 3.0, Sn 3.3, Te 13.8, Cs 6.1 at% Rim,
Pd 70.1, U 3.3, Sn 4.9, Te 20.8, Cs 0.9 at%
None of the inclusions constituents. 4. Theoretical
were found to contain clad
considerations
4. I. Fuel composition The irradiation behaviour of these fuels is dominated by the chemical state of the fission products
Table 3 Composition of fuel after irradiation. Initial composition Uo.asPuo. IsOz+x Pin no. 1
2
3
4
___~ Matrix elements U Pu Y Lanthanides
0.8420 0.0781 0.0015 0.0336
0.8413 0.0773 0.0016 0.0357
Total matrix elements
0.9552
0.9559
0.8424 0.0684 0.0017 0.0391
0.8437 0.0780 0.0016 0.0344
0.9516
0.9577
-____ Oxide formers
Metal
Oxide
Metal
Oxide
Metal
Oxide
Metal
Oxide
CS Ba Zr Sr Nb MO
0.0110 0.0048 0.0151 0.0030 0.0002 0.0174
0.0022 0.0048 0.0302 0.0030 0 0
0.011.5 0.0052 0.0161 0.0032 0.0003 0.0184
0.0058 0.0052 0.0322 0.0032 0.0006 0.0368
0.0125 0.0056 0.0175 0.0035 0.0002 0.0203
0.0025 0.0056 0.0350 0.0035 0.0004 0.0390
0.0112 0.0049 0.0155 0.0031 0.0003 0.0176
0.0022 0.0049 0.0310 0.003 1 0.0006 0.0352
Clad (1) Initial O/M Final O/M (1) Fraction MO oxidised Valency of U Pu + Lan
1.94 1.99 0 4 3.85
(1) Separate calculation for 10 Mm of clad corrosion.
0.08 2.15 2.162 2.078 1.00 4.36 4
2.00
2.023
2.012
2.032
0.96 4.025 4
1.00 4.073 4
263
F.T. Ewart et al. / Composition changes in irradiated oxide fuel materials II 2
0’ in
kc&.
mole-’
2
2 OL OftA
03,
2
I
1
I
.i
2
3
Fuel
radius
2,
:::lYL (
1.I
(cm) O,t
a+=7
1
Fig. 10. Temperature versus fuel radius for each pin at start of life. l!
A theoretical assessment has therefore been undertaken to assist the interpretation of the results. The composition of each of the fuels after irradiation has been calculated using the ~NBRN computer programme [7]. The compositions which were obtained were used for an oxygen balance calculation which is shown in table 3; the oxygen is allocated to the fission products as follows: (a) The stable oxides formed are BaO, ZrOz (as BaZrO,), SrO, NbO. (b) The less stable oxides are formed only above certain U or Pu valencies; NbO, at Pu valency 3.9, Cs,O at valency 3.8, MoOZ at U valency 4.006 and Cs,O at 4.1 (estimated). (c) The matrix cations, U, Pu, Y, lanthanides and residual Zr dissolve in the fluorite lattice and adopt the mean valency of the lattice [ 11. (d) The noble metals form metallic alloys, and the rare gases remain as elements. The molybdenum and technetium which are present in these alloys will exist in equilibrium with their oxides according to the oxygen potential of the environment. generated.
-
I 1 Fuel
radius
I
1
.2
.3
(cm)
Fig. 11. O/M ratio versus fuel radius for each pin at start of life.
Further considerations are necessary for cases involving appreciable divergency from stoichiometry; there are, in pin 1, the formation of the ‘PuPd3’ phase and in pin 2, the oxidation of the clad and insulator pellets. Consider the fission product inclusion in pin 1: This inclusion contains an alloy of palladium with uranium and plutonium, and therefore reduces the heavy metal concentration in the matrix. The extent of the reduction can be estimated on the basis that all the ruthenium generated is present in the noble metal phase associated with a fraction of the palladium. The balance of the palladium not in these inclusions is alloyed with uranium and plutonium. This fraction is 1.8% of the plutonium and 0.2% of the uranium and has been deducted from the uranium and plutonium atomic ratios in table 3.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
11
-50
r
-2‘o-
1500 I
1 Fuel
.2 radius
.3
Icm)
I
/ .l Fuel
radius
I
/
2
3
(cm)
Fig. 12. Oxygen potential versus fuel radius for each pin at start of life.
Fig. 13. Temperature versus fuel radius for each pin at end of life.
Consider next the case of pin 2, which had an initial oxygen to metal ratio of 2.15. This represented a highly oxidising environment, with respect to the clad, at the start of the irradiation. The oxygen balance calculation is, therefore, complicated by the oxidation of the clad and other species. Caesium will be oxidised to Cs,O and molybdenum will be oxidised to MOO, completely. The oxygen potential gradient must, therefore, be reflected in some other equilibrium, for instance Tc/TcO,. Oxidation of the clad is difficult to quantify; for the purpose of discussion, oxidation to a uniform depth of 10 pm is assumed. If the oxidation product is taken to be M,O, then the oxygen requirement is 0.08 atoms per oxygen per mole of fuel, and the resultant oxygen to metal ratio of the fuel under these conditions is given in table 3. Note that this assumption predicts that the fuel has reduced during irradiation. In the case of a highly oxidised fuel, the insulator pellet of UO2 represents a further oxygen sink. The extent of oxygen diffusion into the insulator can be assessed from the ceramographs, assuming that the hemispherical crack repre-
sents the limit of oxygen diffusion. These show that oxygen has diffused to form a 2 mm radius hemisphere, presumably at both ends of the fuel stack. If the uranium oxide increases in oxygen to metal ratio to the original value for the fuel, the oxygen requirement is 0.0045 atoms of oxygen per mole of fuel, which can be considered negligible. This oxygen requirement is likely also to be an overestimate since the equilibrium oxygen to metal ratio will be lower than the average value of the fuel. The. oxygen balance calculations adopting all these assumptions are detailed in table 3. The oxide forming fission products account for a fraction of the original oxygen and the balance, compared with the concentration of the matrix elements, provides the final oxygen to metal ratio. An alternative value is included for pin 2 allowing for clad corrosion. 4.2. Oxygen and temperature distribution The previous section was concerned with the overall change in oxygen-to-metal ratio of the fuel during
265
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsII
2.3 din
02 \ \ \
kcds.mole-' 04
\
\
03
\
2-i,_
\ IO 2.'
O’-
Oh
io-
21o-
L
1.9! j-
I
I
1
1
.2
.3
Fuel
radius
(cm1
Fig. 15. Oxygen potential versus fuel radius for each pin at end of life. 19 O-
I
I
.l
.2
Fuel
radius
I -3
Icml
Fig. 14.0/M ratio versus fuel radius for each pin at end of life.
irradiation. It is now necessary to consider the mutually interacting effects of temperature gradient and oxygen redistribution in that gradient. The redistribution of oxygen in an operating fast reactor has been discussed theoretically in a number of papers, discussed in the review by Bober and Schumacher [2]. These various treatments consider the oxygen distribution as the resultant of the effects of gaseous and solid state transport. This is considered to be represented by the equation: Inx=Q*IRT+c,
(1)
where x is the deviation from stoichiometry, R and T are the gas constant and absolute temperature respectively, c is a constant and Q* is the heat of transport for the oxygen transport process. A limited number of experimental observations of oxygen re-distribution
have been made on actual fuels [8,9] and in laboratory experiments on fuel materials subjected to temperature gradients [ 1O-l 21. These observations have given support over a limited temperature range to the theoretically derived expression for oxygen redistribution. Values of between 5 and 40 kcal/mole for the heat of transport Q* have been found under varying conditions of initial oxygen-to-metal ratio of the fuel. These values have then been used in the fuel chemistry computer modelling programme TPROF [ 131 to calculate the oxygen redistribution and oxygen potential within the fuel pin. This programme models fuel behaviour, it is used here to compute a temperature profile for the fuel and then to distribute the oxygen according to eq. (1). The thermal conductivity is then modified to accord with the calculated distribution; the programme repeats these calculations until satisfactory convergence is obtained. At each time step during the irradiation, the mean oxygen to metal ratio is calculated as outlined in section 4.1, and the temperature and oxygen profile determined. The results of these calculations are given in figs.
F. T. Ewart et al. / Composition
266
changes in irradiated
1000 Temperature
Fig. 16. Ellingham diagrams of
U, Pu
oxide fue! materialsII
1500
2000
(Cl
oxide [ 14,151, some fission product and clad component systems [161.
lo-15 where graphs are drawn of temperature, oxygen to metal ratios and oxygen potentials as a function of fuel radius for the start of life (figs. 10-12) and the end of life (figs. 13-l 5) conditions. The values of Q* selected for the computation of each oxygen-to-metal ratio profile are shown on each curve. The relationship between oxygen-to-metal ratio and oxygen potential used for these calculations is given in fig. 16. The values used are interpolated between those of Markin and McIver [ 141 and Javed [ 1S].
5. Discussion
The oxygen content of the fuel and its redistribution will result in differing conditions of oxygen potential which will control the local chemistry of the fuel and will be reflected in the type and composition of the segregate fission product phases formed. Since is has not been possible to measure local oxygen potentials in the fuel directly, a correlation has been attempted between the theoretical treatments described
and the observations metal inclusions.
of the composition
of the noble
5.1. Experimerztai indications of oxygen migration The oxygen redistribution can be illustrated by the equilibrium states of the oxidation reactions of the intermediate oxides. The two equilibria Mo+O*+ MOO, and Tc + 0, * Tc02 are particularly convenient in that the metals form segregate phases with the noble metals, and the oxides disperse into the matrix. The state of the equilibrium can, therefore, be readily detected with the electron probe micro-analyser. An approach to the quantitative interpretation of such data has been made to this problem previously by Johnson et al. [17]; such an interpretation is not attempted here, principally because of instrumental limitations. In pin 1 the MO and Tc in the metallic inclusions are present in ratios similar to their fission yields; the only indicator of local oxygen potential is the PuPd, phase formed in the fuel rim. In pin 2 there is clearly adequate oxygen available
261
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsII to oxidise the molybdenum
and a significant graction of the technetium also, this is evident from the ‘white’ inclusion composition. The existence of the molybdenum oxide phase in the insulator pellet suggests that the molybdenum has been oxidised to Mo03, a volatile oxide, which has distilled into the colder regions of the fuel. The 0, t 2Mo02 - 2Mo0, reaction is shown in fig. 16 and considering the extent of extrapolation of the thermodynamic data and the oxygen redistribution model to the O/M and temperature involved in this pin, it is not unreasonable to expect this reaction to occur. The data obtained here from pin 3 are in general agreement with the described models, that is, that an increasing fraction of the molybdenum is oxidised as the radius decreases; in pin 2 the system is much more oxidising, all the molybdenum has been oxidised and a similar gradient is observed in the technetium concentration. The observations from pin 4 are anomalous in that there is no detectable molybdenum concentration gradient, which is surprising when the evidence from ceramography was that the two types of pellet had equilibrated. There is, however, sufficient oxygen available in this pin to oxidise all the molybdenum. Some of the molybdenum has been retained in the metallic state, suggesting perhaps that some caesium oxide equilibrium acts as an oxygen sink before all the molybdenum is removed from the noble metal alloy. 5.2. Fission product segregation A large number of different fission product segregate phases have now been reported in the literature and have been reviewed by Kleykamp [ 181. In very few of the published works has it been possible to relate the existence or composition of the phases to the oxygen-to-metal ratio of the fuel or the irradiation conditions. The significance of this series of irradiations is that both the fuel and the irradiation conditions are well defined thus providing an opportunity to indicate the conditions of formation of a number of inclusion types. Table 4 lists the inclusions found and the estimated local oxygen potential deduced from fig. 16. The PuPd, type of inclusion found in pin 1 has been previously reported by Bramman [ 19 ] and Kleykamp [ 181 in sodium-logged failed fuel pins. These provide a relatively reducing environment which is consistent
Table 4 Fission product inclusions Pin no.
Inclusions
Temp (OC)
1
UtsPurrPd74 MoaTTctoRwRhrtPds P_ddUPu
1150 1700 700
Estimated oxygen potential (kcal/mole) -132 -125 -130
1950 T~~Rtre~Rh~~Pd~ 1950 PdwTeaaU4 1950 cs (01) 650 Tc MO (O?)* 650 Fe Cr Cs MO 6.50 FeCr Ni U Pu Cs Ba 650 pd Te Cs (U, Pu?) 650 PdeTesU *Volatile species originally in matrix
+ 20 + 20 + 20 -100 -100 -100 -100 -100
FeCr Ni Cs FeM &&Tc&R_hPd
900 1100
- 95 - 95
Pd74UaSnsTet4Cse Pd7eUsSnsTearCsr MO - Tc Ru - Rh Pd
1750 1000 general
2
3
4
general - 20 - 65 >- 80
Where analyses are not given, the principal constituents underlined.
are
with the oxygen potential calculated by Holleck [20] to be necessary for the formation of the URu3 , URh3 phases. The oxygen potential calculated was -125 kcal/mole, which shows some agreement with the value derived here (-130 kcal/mole) for the rim of the pin 1 fuel. The complex palladium-tellurium-caesium phases found in pins 2 and 4 are more unusual in that there is only one published reference to these phases (Huber and Kleykamp [21]). These workers found Pd, Te, Sn alloys in the columnar grains of 15% mixed oxide at an initial O/M of 1.98 which had reached 6.5% burnup. This fuel, which contained 11.5% 235U would have its chemistry controlled by the MO/MOO, equilibrium. The observation is in contrast to the present one where Pd, Te phases occur in the fuels which have exceeded this control. 5.3. Fuel-cladding interactions Although this experiment consisted of single pin irradiations, the results relate well to the experience
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materialsII
268
Table 5 Clad corrosion products ~__ ~~__ Pin no.
Inclusions
Oxygen potential (kcal/mole)
1 2 3 a) b) 4
CS 9 Ba Fe Cr Ni QFeCrG FePd Cs -_ Ba Te Nd MO
-120 - 90 ~ 90 - 90 - 90
Principal constituents
has been observed also by Kleykamp [ 181, no cladding elements were found in the fuel matrix remote from the clad. The Argonne work [25], for example, reports up to 13% iron in the noble metal inclusions and Crouthamel [ 171 has reported large numbers of iron rich inclusions deep in the fuel. The German work, however [18,21], reports Fe, Pd and Fe, Ni, Pd alloys near the fuel surface and in one case [ 181 reports a clad interaction similar to that found in pin 2, but occurring in a fuel pin of starting O/M 1.98 f 0.01.
are underlined.
5.4. Heavy metal concentrations found in large-scale irradiation experiments. The modes of attack, intergranular and broad front have been reported previously 121-231 in considerable detail. The results from these experiments support the belief that clad attack can be minimised by low oxygen potential conditions at the clad and may have some bearing on defining the conditions that give rise to intergranular or broad front attack. Table 5 summarises the analyses of the reaction products in the fuel-clad gaps. The predominant species found in these pins was caesium. In pins 2 and 3, where the final oxygen to metal ratio had exceeded the control offered by the MO/MOO, equilibrium, barium was also present. The observations on pin 1 suggest that caesium is compatible with the clad at a local oxygen potential of - 130 kcal/mole. The difference in the rim oxygen potentials of the other pins (-~90 to -9.5 kcal/mole) are small and do not account for the differences in corrosion mechanisms. However, the enrichment of chromium in the reaction product in pin 2 may be significant, in that interaction of caesium oxide and chromium could remove the protective oxide layer from the steel and would allow a ‘broad front’ attack. There is evidence [24] that the Te : Cs ratio may change the oxygen potential at which a particular mode of clad attack operates. The differences in corrosion phenomena found here in fuels pins having apparently similar oxygen potentials at the clad may be accounted for by the widely differing oxygen potentials in the body of the fuel which could control the relative rates of release of tellurium and caesium to the fuel-clad gap. A significant difference between these irradiations and those reported by other workers [20,22,25] is in the fission product-clad component phases. With the exception of the Fe, Pd phase found in pin 3, which
Some consideration must be given to the composition of the matrix in Pu-enriched fuels which are irradiated in thermal fluxes to high burn-up. The assumption is made, during discussions of fast reactor fuel, that the chemistry of the matrix is controlled by the valency state of the uranium (O/M > 2) or plutonium (O/M < 2). In thermal irradiations, however, considerable Pu burn-out occurs and the lanthanide concentration in the matrix becomes a significant (fig. 17) fraction of the total presenting a differing chemical situation from that in a fast reactor. This effect is demonstrated in fig. 17. For an MTR irradiation, the majority of fissions are in plutonium and plutonium replenishment by neutron capture is relatively unimportant. This is compared with a Dounreay Fast Reactor (DFR) irradiation where fissions are in both uranium and plutonium and with a Prototype Fast Reactor (PFR) irradiation where fissions are in plutonium and there is a significant Pu replenishment rate. The chemistry of the resulting system is thus complex and is controlled by the solid solution of U, Pu and lanthanides. There are very few thermodynamic data available for solutions of lanthanides in Urania, plutonia or mixed systems. The observations of Markin and Crouch [26] show that the oxygen potential of ceria-Urania solution is higher than that of equivalent plutonia-Urania solutions in the hypostoichiometric regions; the observations of Stadlbauer [27] in the hyperstoichiometric region of lanthania-Urania are in agreement with this trend. At this stage, the available data are inadequate to allow a quantitative assessment of the effects of Pu burn-out. There is, however, a reasonable indication that the oxygen potential of the oxide matrix will be,
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsIf Information corrosion
DFR 15%Pu 79 % 235u
269
on the cause and mechanism
is difficult
to achieve
of clad
on the basis of single
pin experiments. It can be concluded, however, that an oxygen potential at the clad of less than -130 kcal/mole is likely to inhibit clad corrosion. It is also apparent that the distinction between intergranular and broad front corrosion, both of which occur at -90 to -95 kcal/mole, is not simply attributable to the oxygen potential at the surface of the fuel but is a function of the chemical species present in the fuel/ clad gap. These will depend on the transport mechanisms available and the oxygen potential elsewhere in the fuel.
Lonthonsde
Acknowledgements MTR 15% Pu
Fig. 17. Concentration of piutonium and lanthanide elements as a function of burn-up.
in the case of fuel irradiated in thermal fluxes, significantly more oxidising than the equivalent fast reactor irradiation because of the higher rate of plutonium burn-out.
Acknowledgement must be made to the considerable skills in sample preparation shown by the Process Workers of Harwell without whom this work would not be possible. The experimental assistance with the EPMA of Mr. L. Murphy is also acknowledged. Dr. F. Caligara of Euratom, Karlsruhe is thanked for allowing the use of his computer code ‘TPROF’ for fuel performance modelhng.
References 6. Conclusions The general trend of observations made on the chemical form of the fission products is in agreement with a model of increasing oxygen-to-metal ratio with increasing burn-up. The variations in composition of the noble metal inclusions with fuel radius conform in general with theories of oxygen redistribution and the particular observations of the PuPd, phase are in good agreement with the model of oxygen redistribution used. In order to obtain quantitative data relating to oxygen redistribution and test the validity of the model, there is a need for improved experimental techniques of oxygen potential measurements. Development of techniques for measurement is of ~portance, but a deal of information would be available if the conditions of formation of the various fission product segregate phases were known.
[l] J.H. Davies and F.T. Ewart, J. Nucl. Mat. 41 (1971) 143. [ 21 M. Bober and G. Schumacher, Adv. in Nucl. Sci. and Tech. 7 (1973) 21. [3] J. Bramman and H.J. Powell, J. Brit. Nucl. En. Sot. 14 (1975) 63. [4] W. Batey and K.Q. Bagley, J. Brit. Nucl. En. Sot. 13 (1974) 49. [ 51 J. Adam and F.T. Ewart, BNES Symposium on Post-irradiation Examination Techniques, Reading f1972). [61 D.M. Poole, UKAEA Report AERE-2469 (1971). [ 71 F.T. Ewart and H. Kaneko, UKAEA Report AERE R7191(1972). [8] C.E. Johnson et al., Reactor Technology 15 (1972-73) 303. [9] E.G. Aitken, S.K. Evans, M.G. Adamson and J.H. Davies, IAEA Symposium on Thermodynamics of Nuclear Materials, Vienna (1974). [lo] A.E. Aitken et al., J. Nucl. Mat. 30 (1969) 57. [ I1 ] M.G. Adamson and R.F.A. Carney, UKAEA Report, AERE R-6831 (1972). [ 121 M.G. Adamson and R.F.A. Carney, UKAEA Report, AERE R-6830 (1972).
270
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials I1
[ 131 F. Caligara, Euratom, Karlsruhe, private communication
[ 141 [ 151 [ 161
[ 171 [ 181
[ 191 [20]
(1975). T.L. Markin and E.J. Mclver, Plutonium 1965 (Chapman and Hall, London 1967) p. 845. N.A. Javed and J.T.A. Roberts, USAEC Report ANL 7901 (1972). 0. Kubachewski, E. Ll. Evans and C.B. Alcock, Metallurgical Thermochemistry, 4th ed. (Pergamon Press, London 1967). C.E. Johnson and C.F. Crouthamel, J. Nucl. Mat. 34 (1970) 101 H. Kleykamp, IAEA Panel on the chemical state and fission product behaviour in irradiated nuclear fuels, August 1972. J.I. Bramman, R.M. Sharpe, D. Thorn and G. Yates, J. Nucl. Mat. 25 (1968) 207. H. Holleck and H. Kleykamp, GfK Report, KFK-1181 (1970).
[21] H. Huber and H. Kleykamp, GfK Report, KFK-1324 (1972). [22] R.B. Fitts, E.L. Long and J.M. Leitnaker, ANS Conf. on Fast reactor fuel element techno’--y, New Orleans, April 1971 (Am. Nucl. Sot. 1971) p. 431. [ 231 K.J. Perry, G.F. Melde, W.H. McCarthy and R.N. Duncan, ibid p. 411. [ 241 M.G. Adamson et al., IAEA Symposium on Thermodynamics of Nuclear Materials Vienna (1974). [25] Argonne National Laboratory, Chemical Engineering Division Research Highlights USAEC report ANL 7750 (1970). [26] T.L. Markin and E.L. Crouch, J. Inorg. Chem. 32 (1970) 77. 1271 E. Stadlbauer, GfK Report KFK-1649 (1972).
254-270
THE CHEMICAL EFFECTS OF COMPOSITION CHANGES IN IRRADIATED OXIDE FUEL MATERIALS II-FISSION PRODUCT SEGREGATION AND CHEMICAL EQUILIBRIA F.T. EWART and R.G. TAYLOR UKAEA, Applied Chemistry Division, Atomic Energy Research Establishment. HarweU, Oxfordshire OX1 1 ORA, lJK
J.M. HORSPOOL and G. JAMES UKAEA, Research and Development Laboratories, Windscale Works, Sellafield, Cumbria CA20 IPG, UK Received
9 February
1976
Four miniature fuel pins containing uranium/plutonium oxide, covering a wide range of oxygen-to-metal ratios, have been irradiated in a thermal reactor. These pins were subjected to a detailed examination, by electron probe microanalysis and ceramography. The results of this analysis have been interpreted in the light of the results of a computer model of the pin behaviour . The widely different fuel chemistry in each of the fuel pins produced different fission-product phases and different clad corrosion phenomena. The phases found are in general agreement with the model which predicts increasing oxygen-to-metal ratio with increasing burn-up and oxygen redistribution in the temperature gradient. Clad corrosion is seen to be inhibited at low oxygen potentials but the distinction between intergranular and ‘broad front’ corrosion is not attributable
simply
to the local oxygen
potential.
Quatrc epingles de combustible miniatures contenant de l’oxyde mixte d’uranium et dc plutonium, couvrant un large domaine de rapport oxygene/metal, ont Cte irradiees dans un reacteur thermique. Ces dpingles furent soumises i un examen detail16 par microsonde Clectronique et ceramographie. Les resultats de cette analyse ont dtd interprdtes a la lumierc dcs rcsultats d’un modele de calcul sur lc comportement des aiguilles de combustible. La chimic du combustible tres differentc pour chacune des aiguilles combustibles avait pour consequence de produirc diffdrcntcs phases de produits de fission et divers phenomenes de corrosion du gainage. Les phases trouvecs sont en general d’accord avcc le modtle qui prdvoit un accroissement du rapport oxygene/metal avec le taux de combustion et la redistribution de I’oxygenc dans le gradient de temperatures. La corrosion du gainage apparait &tre inhibee aux faibles potent& d’oxygene, mais la distinction entre corrosion intergranulaire et ‘front general’ n’est pas attribuable simplement au potentiel total en oxygene. Vier kurze Uran-Plutonium-Oxid-Brennstabc mit schr unterschiedlichem Sauerstoff/Metall-Verhahnis uurden in cinem thcrmischcn Reaktor bestrahlt. Die Brennstabe wurden keramographisch und mit der Elektronenstrahlmikrosondc eingehend untersucht. Die Ergebnisse diescr Analyse wurden unter dcm Aspckt der Ergebnisse eincs Rcchenmodells fir das Brennstabvcrhalten gedeutet. Die sehr unterschiedliche Chemie des Brennstoffs in jcdem der Brennstabe hat unterschiedlithe Spaltproduktphasen und Htillkorrosionsphanomene zur Folge. Die beobachteten Phasen stehen in ungefahrer Ubereinstimmung mit dem Modell, das tin steigendes Sauerstoff/Metall-Vcrhaltnis mit dem Abbrand und tine Sauerstoff-Umverteilung im Temperaturgradicntcn voraussagt. Die Hiillkorrosion ist bei niedrigen Sauerstoff-Potentialen gehemmt, der Unterschied zwischcn Korngrenzenund flachcnhafter Korrosion kann nicht einfach auf das lokale Sauerstoff-Potential zurtickgeftihrt werden.
fuel by a range of fission product species were discussed and the likely chemical form of the principal fission products was considered. Supporting experimental evidence was presented and it was concluded that there are two principal effects: firstly, that the average valency of the fission products is less than that
1. Introduction In an earlier paper [ 11, experimental evidence was presented which led to conclusions about the chemical state of an irradiated oxide fuel. Chemical modifications caused by the replacement of fissile atoms in the 254
I? T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
of the fissile species such that the fuel becomes more oxidising with burn-up; and secondly, that some of the fission products will form separate phases which affect the overall oxygen balance and may modify the performance of the oxide as a reactor fuel. An additional effect that was not considered in these earlier experiments, which were under near isothermal conditions, was the extremely steep temperature gradient which obtains in an operating fast reactor type fuel pin (about 800 K/mm). The principal effect of this gradient in the oxide fuel system is to provide the driving force for oxygen redistribution. This, in turn, has an important bearing on the operational performance of a fast reactor fuel pin and the optimum choice of oxygen-to-metal ratio for the initial fuel. The magnitude of oxygen redistribution effects has been the subject of a number of theoretical assessments. A few experimental measurements have also been made; both of these aspects have been reviewed by Bober and Schumacher [2]. Experimental work on the phases formed in fuel during irradiation is described by Bramman and Powell [3]. Results are presented here from the irradiation of four miniature fuel pins which were conducted under carefully controlled conditions and then subjected to a detailed post-irradiation examination using electron probe analysis and other techniques. The initial oxygen-to-metal ratio of these pins was selected to cover a wide range, from O/M = 1.94 to O/M = 2.15. This provided an opportunity to relate the previous theoretical and experimental studies to actual fuel pin irradiations and to illustrate the effect of the thermochemical changes and the differences in fuel behaviour relating to the different conditions of oxygen-to-metal ratio. The limits of desirable oxygen-to-metal ratio for an irradiated fuel system are set in the upper range by compatibility with the stainless steel [4] clad and, in the lower range, by the formation of free metal in the fuel. Under isothermal conditions the range between these limits is large but in the operating fuel the oxygen redistribution in the temperature gradient and the change in oxygen-to-metal ratio during burn-up may make it impossible for the fuel not to exceed these limits during its operational life, depending on the magnitude and direction of the redistribution of oxygen and the rate of change of the oxygen-to-metal ratio.
255
The behaviour of the fuel pins under these extremes of initial fuel specification has been followed principally by the identification of fission product phases and other structural features observed in the fuel. These observations have been related to the local chemical conditions within the fuel, using a computer model describing the chemical state of the fuel. Whilst uncertainties exist both in definition of conditions and the mechanisms controlling the behaviour, it is clear that widely differing chemical conditions existed during the irradiations which were reflected in the phases developed within the fuel.
2. Experimental
procedure
The fuel within the four fuel pins for irradiation was derived from two batches of powder manufactured by a common route. The powders were then compacted and sintered using identical methods and then adjusted to three different oxygen-to-metal ratios. Each fuel pin contained approximately 12 grams of (U, 85Pu0 15) oxide to the following compositions: Pin 1 O/M = 1.94; carbon 1550 ppm Pin 2 O/M = 2.15; carbon
Details of the fuel are given in table 1, the irradiation conditions are in table 2. Each pin was 175 mm long, 5.84 mm outside diameter and 5.08 mm inside diameter. The clad was 20% cold-worked M3 16 stainless steel. The fuel stack was contained between two UO2 pellets and was retained by a collet, leaving a gas plenum of 70 mm. The irradiations were carried out in a static sodium-filled capsule for 16 reactor cycles in the PLUTO Materials Testing Reactor at Harwell. The nominal fuel rating was 200 W/g and the clad mid-wall temperature was 650°C. The samples of fuel were taken for electron probe micro-analysis (EPMA) from positions in the fuel pins shown in the cutting diagram in fig. 1. The transverse sections were mounted in conduct-
256
I;. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
Table 1 Characterisation
of fuel before
irradiation 99312
993/l Pu content Isotopic ratios
&
15.2
15.23
15.23
~238
0.73 99.27
0.03 0.89 99.08
0.03 0.89 99.08
Pll239 Pu2‘+0 PUZ4’
93.54 6.00 0.46
93.43 6.08 0.49
93.43 6.08 0.49
(
U
23h
) ’
-
u235
1555 < 100 170 400 260 80 < 100 660 < 100 < 10 -
Carbon cu Ca Al Mg Cr Mn Fe Ni Cd Zn O/M Sintering
99313
1.94 1 hr/1550 C/H2
conditions
Density % Theoretical
< 50 100 140
epoxy resin in a short bakelite cylinder and reduced in thickness to about 125 Mm thick, when the surface was polished to a 1 pm finish. Throughout these operations, nonpolar solvents were used to pre-
C/CO2
2.000 3 hr/1600 C/H2
99.2
serve the fission product caesium. At this thickness the radiation levels from the samples were sufficiently reduced to be acceptable to the microprobe analyser. The longitudinal samples were taken from the sam-
ing
Table 2 Details of irradiation
Two pellets of 993/2 in centre of stack of 993/3
50
98.2
95
99314
conditions Pin no.
Mean rating (w/g) Linear rating (kW/m) Irradiation time (MW/h) Average flux (n/cm’ . s) Can nominal midwall temp (“C) No. of reactor cycles Burn-up (%) Gas release (%)
1
2
3
4
165 29.0 196704 3.28 x 1Or3 650 16 7.3 63.7
183 33.5 196252 3.82 x 1013 650 16 8.1 71.5
199 36.4 200577 4.24 x 1013 650 16 9.2 82.1
172 31.8 196668 3.41 x 10’3 650 16 7.6 79.1
251
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materials11 993/L/01 9931212A
9931112A 993/3/2A
I
I
End
cop
Insulator
I
High O/M pellets in 9931.L
Approx
993/1/1A \993/2/1A
‘=/3/1A, I
I
Insulator
10 cm
Fig. 1. Diagrammatic section of fuel pin showing sections removed for analysis.
ples mounted for ceramography. These had been obtained by reducing a cylinder of fuel to a semi-cylinder. A slice 2 mm thick and 6 mm square was taken from the face of the ceramographic sample and was mounted and ground as previously. The bulk removal of material was from the rear of the sample so that the face finally observed was within the 125 pm of the original. The electron probe micro-analysis was carried out on a Cameca MS85 and a Cambridge Microscan 5 instrument. The development and shielding of these instruments has been described elsewhere [5]. The detailed searching of the sample surface for fission product segregation was carried out on the Cameca instrument which had a detection sensitivity under search conditions of 0.5 wt% of noble metals or clad constituents and 1 wt% of rare earths. Under the conditions used, the limit of detection when analysing on a fixed point was 0.1 wt%. It has been found that with these instruments it is not possible to obtain quantitative data from inclusions smaller than a 5 pm diameter hemisphere, because of errors introduced from matrix contributions due to the spread of the electron beam. However, relative concentrations may be estimated with reasonable accuracy for smaller inclusions. Where these are presented in the results, the ratios are those of the Xray intensities, which are approximately proportional to the weight fractions of the element. The quantitative data presented have been collected from uniform, dense, polished areas exceeding 5 pm in diameter. The data were reduced and corrected using the computer programme DMPEMP8T [6]. This programme presents the results as weight percentages and subsequently as atomic percentages normalised to 100%. Results were accepted for normalisation only when the range of the weight percentages was between
98% and 102%. Analyses falling outside this range were repeated under improved conditions or further elements were sought.
3. Results 3.1. Electron probe analyser observations 3.1.1. Pin 1 Two sections were examined from this pin, the transverse section 993/1/2A was taken from the centre of the fuel stack and the longitudinal section 993/ l/ 1A was taken to include the upper (plenum end) insulator pellet. The transverse section (fig. 2) contained two dif-
Fig. 2. Transverse section of pin 1.
X
12.5.
258
F.T. Ewart et al. f Composition changes in irradiated oxide fuel materials II
ferent metallic inclusions. The well known noble metal inclusion was found in the central void; this had the composition: MO 37.5, Tc 9.8, Ru 36.7, Rh 11.1; Pd 4.9 at%. Inclusions of similar qualitative analysis were found in the columnar grain region. A different type of metallic inclusion was found in a number of positions about 120 pm from the outer edge of the fuel (fig. 3). These were of a diffuse form and contained U, Pu and Pd in the approximate composition (U, Pu)Pd,. The composition was not, however, constant and some analyses obtained were: U 15.3,Pu
11.4,Pd73.3
at%
U 15.6, Pu 13.2, Pd 71.2 U 17.8, Pu
7.7, Pd 74.5
The fuel was examined for concentrations of clad species but norie was found. The fuel clad gap contained some well defined deposits which were found to contain only caesium. Caesium was also found in the fuel rim but was not found in the clad. The longitudinal section showed the presence of apparently metallic particles at the interface between the insulator pellet and the fuel stack. These inclusions contained U, Pu, Pd, Sn and the composition varied along the fuel radius. Approximate compositions (*5%
Fig. 3. PuPd3 type inclusion
pin 1. X 80.
on each component)
in atomic percentages
were:
Outer U1S,6Pd67.5Sn16.9 at% Inner Ug,3Pu4.1 Pdj8 3snS1 3 at% . 3.1.2. Pin 2 Two sections were examined from this pin, a transverse section 993/2/2A taken from position shown in fig. 1 and a longitudinal section 993/2/1A inciuding the plenum and insulator pellet. A micrograph of the longitudinal section is shown in fig. 4. The prominent features of both these sections were the extensive clad corrosion (fig. 5) and, in the longitudinal section (fig. 4), the complex group of phases found at the insulator-fuel interface. The observations will be presented in three groups, fuel matrix, insulator interface, and clad interface. Fuel matrix The surface microstructure features of the fuel matrix were similar to those commonly found in high burn-up oxide irradiations. White metallic inclusions
Pig. 4. Fuel-insulator x 12.5.
region,
longitudinal
section
pin 2.
F. T. Ewart et al, / Composition changes in irradiated oxide fuel mare&Is II
FULL Fig. 5. Clad corrosion
pin 2. x 280.
were found in the equiaxed and columnar grain region. A second type of metallic inclusion, darker than the first, was found in the fuel rim. The white inclusions were shown to be the expected noble metal inclusion but molybdenum was found to be absent; typical compositions for inclusions in the fuel centre position were: Tc 8.5, Ru 63.1, Rh 21.0, Pd 7.4 at% Tc 8.8, Ru 62.5, Rh 19.8, Pd 8.9 at% The Tc : Ru ratio in the inclusions was measured at several points across the fuel radius, and was found to be 1: 7 in the central void 1 : 11 in the columnar grains 1 : 7 in the equiaxed grains and 1 : 4 in the rim. The fission-product molybdenum was found only in the fuel matrix. The darker inclusions which occurred up to 200 pm from the clad contained U, Pd, Te in the approximate atomic ratio 1 : 6 : 3. Insulator interface
The area shown in fig. 4 is a longitudinal section of the interface between the fuel and the UO, upper in-
259
sulator pellet. The dominant features of this section are the hemispherical crack, up to 0.5 mm wide, which had formed in the U02 insulator pellet and the apparent extrusion of the insulator into the central void of the fuel stack. This section was examined by arautoradiography and EPMA to attempt to identify the causes of the hemispherical crack. No significant diffusion of plutonium into the insulator could be found and the penetration of fission products into the insulator was bounded by the crack. It is postulated, then, that the crack represents the limit of oxygen diffusion into the insulator or a temperature boundary. The material in the central void consisted partly of the extruded insulator and partly of the fuel forming a bridge. The section contained a large number of separate phases. The bridge contained a large inclusion, some 300 E.trnin diameter, which had the composition Pd 73.7, U 4.0, Te 22.3 at%. Smaller inclusions of similar composition were found in the fuel rim both in this section and in the transverse section of this pin. In this region, these inclusions were often associated with noble metal inclusions and a darker phase which :ontained only caesium, presumably as the oxide, as hewn in fig. 6. Where the extruded insulator pellet met the fuel bridge an iron and chromium-bearing phase containing only traces of nickel was found. A further phase was found in the insulator pellet about 1 mm below the hemispherical crack. The only metals found were MO and Tc at 10 and 57 wt% respectively, leading to the conclusion that the phase was non-metallic, the most likely form being a mixed oxide phase. About 100 pm from the crack and within 200 m of the clad a non-conducting, comparatively lowmelting phase was found. These properties were shown by the charging of the sample surface and the local melting which occurred when the electron beam was applied to this area. This phase, which contained Cs, Cr, Fe and MO, was thought to be a caesium chromatelmolybdate. In the same radial position as the previous phase, the hemispherical crack contained a complex structure of grain boundary phases which were too small for unequivocal analysis. It could be shown that the grain boundaries contained Fe, Cr and Ni concentrations, however, and were visually similar to the phase found at the clad interface.
260
F. T. Ewart et al. 1 Composition changes in irradiated oxide fuel materials II
which was rich in caesium. In some places on the section this phase was in the ‘gap’, in others it merged into the next layer. (c) A layer of corrosion product up to 75 pm wide. This layer contained Fe, Cr and Ni, the chromium fraction being greater, by a factor of about two, than in the clad itself. To a lesser extent, U, Pu, and Cs, Ba were also found. (d) A complex zone containing a number of bubbles and bounded on the inner side by a row of metallic inclusions. The zone consisted of alternate layers of zone (c) and of fuel. The metallic inclusions were difficult to analyse quantitatively, being less than 5 I.crnin diameter, but were found to contain Pd, Te and some Cs. (e) Substantially unaltered fuel which had in a few places been penetrated by phase (c) to a depth of 100 pm.
Fig. 6. Fission product inclusions pin 2 Pd-Te (light), Metals (bright) and Cs (dark). X 350.
Noble
Fuel-clad interface The clad corrosion on this pin was clearly of the ‘broad front’ type. The maximum clad penetration found was 75 pm; this was observed generally in the longitudinal section and more locally in the transverse section. The reaction product had a slightly different appearance in the two sections. The longitudinal section showed a reaction product attached to the clad and a reaction product on the fuel side of the gap. The transverse section, however (fig. S), showed the clad in contact with the reaction products, but in other places there was a gap between the clad and the reaction products. The examination of these features by the electron probe showed that in all cases the reaction products were the same. These are conveniently described sequentially from the clad to the fuel: (a) Undisturbed clad showing no variations in composition from outer to inner wall and no evidence of second phases in the grain boundaries. (b) A narrow, 10 ym, layer of variable density
3.1.3. Pin 3 Again both transverse 993/3/2A and longitudinal sections 993/3/1A were taken from this pin. The clad in both cases showed pronounced clad attack; this was quite different from the broad front attack seen in pin 2. Almost all regions where the clad was in contact with the fuel showed attack by an intergranular
Fig. 7. Clad corrosion
pin 3.
X
280.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
261
cant decrease in molybdenum content of the inclusions towards the centre of the pin. The values obtained were: Void edge, MO : Ru = 1 : 4
JNCLtiSlON
Fig. 8. Metallic inclusions in pellet interface pin 3. X 280. penetration mechanism (fig. 7). There was no evidence of the complex fuel-clad component phase found in the previous pin, although a metallic phase was observed in a few places at the pellet interfaces (fig. 8). The fuel was of normal appearance with metallic inclusions present in the columnar and the equiaxed grain regions. The elements present in the grain boundary phase in the clad were found to be caesium and the cladding constituents Fe, Cr and Ni. The caesium concentration was reasonably constant throughout this phase whereas the relative atomic concentration of the clad constituents varied widely, e.g. Fe 6-17, Cr 0.8-1.1, Ni 1. There was no variation in the concentrations of the clad constituents across sections of sound cladding. The apparently metallic phase found in the interpellet gaps (fig. 8) was shown to contain only Pd and Fe. In the sample presented it was not possible to carry out an absolute analysis, but the relative intensities corresponded to a composition between FePd and Fe,Pd. Inclusions containing clad components were found in no other locations in the fuel. The metallic inclusions found in the fuel were solely of the noble metal type. None of these inclusions was large enough for quantitative analysis. It was possible, however, to measure the relative concentration of molybdenum and ruthenium. There was a signifi-
Columnar,
MO : Ru = 1 : 3
Equi~ed,
MO : Ru = 1 : 1.8
Rim,
Mo:Ru=
1: 1.4
3.1.4. Pin 4 One transverse section 993/4/Dl for EPMA was taken from a position in the stack of originally stoichiometric pellets. The ceramography of a longitudinal section covering the region where pellets of differing oxygen to metal ratios were located suggested that equilibration had occurred between the two pellets, since the microstructures on each side of the pellet interfaces were identical. This was in contrast to the interface between the fuel and the insulator in pin 2. No evidence was seen on the section taken for EPMA of any clad attack; a slight intergranular penetration was found on other sections (fig. 9). The section contained a deposit in the fuel-clad gap. The fuel contained a number of metallic inclusions; in addition
Y
Fig. 9. Clad corrosion pin 4. x 350.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
262
to the bright noble metal type, a darker type of inclusion was found. Both types of inclusion were widespread in the fuel. As was expected from the visual examination, the electron probe did not detect any clad penetration by foreign elements and there was no evidence of composition variations across the clad. The deposit in the fuel-clad gap was of indefinite density and could not, therefore, be fully analysed; it contained predominantly caesium and barium with lower amounts of molybdenum and cerium, with traces of tellurium and neodymium. The two types of metallic inclusions found in the matrix were analysed by EPMA. The small bright inclusions found in the columnar and equiaxed grain region were the conventional noble metal type containing MO, Tc, Ru, Rh and Pd. As found in pin 3, these were too small for full analysis but the MO : Tc : Ru atomic ratio was measured as a function of radial position:
Void edge, MO : Tc : Ru = 1 : 1.4 : 6.0 Columnar,
MO : Tc : Ru = 1 : 1 .l : 4.2
Equiaxed,
MO : Tc : Ru = 1 : 1.7 : 9.3
Palladium was also found in the darker inclusions which were much larger than the noble metal inclusions and could be analysed quantitatively. Analyses of these inclusions taken at central and rim positions were: Centre, Pd 73.8, U 3.0, Sn 3.3, Te 13.8, Cs 6.1 at% Rim,
Pd 70.1, U 3.3, Sn 4.9, Te 20.8, Cs 0.9 at%
None of the inclusions constituents. 4. Theoretical
were found to contain clad
considerations
4. I. Fuel composition The irradiation behaviour of these fuels is dominated by the chemical state of the fission products
Table 3 Composition of fuel after irradiation. Initial composition Uo.asPuo. IsOz+x Pin no. 1
2
3
4
___~ Matrix elements U Pu Y Lanthanides
0.8420 0.0781 0.0015 0.0336
0.8413 0.0773 0.0016 0.0357
Total matrix elements
0.9552
0.9559
0.8424 0.0684 0.0017 0.0391
0.8437 0.0780 0.0016 0.0344
0.9516
0.9577
-____ Oxide formers
Metal
Oxide
Metal
Oxide
Metal
Oxide
Metal
Oxide
CS Ba Zr Sr Nb MO
0.0110 0.0048 0.0151 0.0030 0.0002 0.0174
0.0022 0.0048 0.0302 0.0030 0 0
0.011.5 0.0052 0.0161 0.0032 0.0003 0.0184
0.0058 0.0052 0.0322 0.0032 0.0006 0.0368
0.0125 0.0056 0.0175 0.0035 0.0002 0.0203
0.0025 0.0056 0.0350 0.0035 0.0004 0.0390
0.0112 0.0049 0.0155 0.0031 0.0003 0.0176
0.0022 0.0049 0.0310 0.003 1 0.0006 0.0352
Clad (1) Initial O/M Final O/M (1) Fraction MO oxidised Valency of U Pu + Lan
1.94 1.99 0 4 3.85
(1) Separate calculation for 10 Mm of clad corrosion.
0.08 2.15 2.162 2.078 1.00 4.36 4
2.00
2.023
2.012
2.032
0.96 4.025 4
1.00 4.073 4
263
F.T. Ewart et al. / Composition changes in irradiated oxide fuel materials II 2
0’ in
kc&.
mole-’
2
2 OL OftA
03,
2
I
1
I
.i
2
3
Fuel
radius
2,
:::lYL (
1.I
(cm) O,t
a+=7
1
Fig. 10. Temperature versus fuel radius for each pin at start of life. l!
A theoretical assessment has therefore been undertaken to assist the interpretation of the results. The composition of each of the fuels after irradiation has been calculated using the ~NBRN computer programme [7]. The compositions which were obtained were used for an oxygen balance calculation which is shown in table 3; the oxygen is allocated to the fission products as follows: (a) The stable oxides formed are BaO, ZrOz (as BaZrO,), SrO, NbO. (b) The less stable oxides are formed only above certain U or Pu valencies; NbO, at Pu valency 3.9, Cs,O at valency 3.8, MoOZ at U valency 4.006 and Cs,O at 4.1 (estimated). (c) The matrix cations, U, Pu, Y, lanthanides and residual Zr dissolve in the fluorite lattice and adopt the mean valency of the lattice [ 11. (d) The noble metals form metallic alloys, and the rare gases remain as elements. The molybdenum and technetium which are present in these alloys will exist in equilibrium with their oxides according to the oxygen potential of the environment. generated.
-
I 1 Fuel
radius
I
1
.2
.3
(cm)
Fig. 11. O/M ratio versus fuel radius for each pin at start of life.
Further considerations are necessary for cases involving appreciable divergency from stoichiometry; there are, in pin 1, the formation of the ‘PuPd3’ phase and in pin 2, the oxidation of the clad and insulator pellets. Consider the fission product inclusion in pin 1: This inclusion contains an alloy of palladium with uranium and plutonium, and therefore reduces the heavy metal concentration in the matrix. The extent of the reduction can be estimated on the basis that all the ruthenium generated is present in the noble metal phase associated with a fraction of the palladium. The balance of the palladium not in these inclusions is alloyed with uranium and plutonium. This fraction is 1.8% of the plutonium and 0.2% of the uranium and has been deducted from the uranium and plutonium atomic ratios in table 3.
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials II
11
-50
r
-2‘o-
1500 I
1 Fuel
.2 radius
.3
Icm)
I
/ .l Fuel
radius
I
/
2
3
(cm)
Fig. 12. Oxygen potential versus fuel radius for each pin at start of life.
Fig. 13. Temperature versus fuel radius for each pin at end of life.
Consider next the case of pin 2, which had an initial oxygen to metal ratio of 2.15. This represented a highly oxidising environment, with respect to the clad, at the start of the irradiation. The oxygen balance calculation is, therefore, complicated by the oxidation of the clad and other species. Caesium will be oxidised to Cs,O and molybdenum will be oxidised to MOO, completely. The oxygen potential gradient must, therefore, be reflected in some other equilibrium, for instance Tc/TcO,. Oxidation of the clad is difficult to quantify; for the purpose of discussion, oxidation to a uniform depth of 10 pm is assumed. If the oxidation product is taken to be M,O, then the oxygen requirement is 0.08 atoms per oxygen per mole of fuel, and the resultant oxygen to metal ratio of the fuel under these conditions is given in table 3. Note that this assumption predicts that the fuel has reduced during irradiation. In the case of a highly oxidised fuel, the insulator pellet of UO2 represents a further oxygen sink. The extent of oxygen diffusion into the insulator can be assessed from the ceramographs, assuming that the hemispherical crack repre-
sents the limit of oxygen diffusion. These show that oxygen has diffused to form a 2 mm radius hemisphere, presumably at both ends of the fuel stack. If the uranium oxide increases in oxygen to metal ratio to the original value for the fuel, the oxygen requirement is 0.0045 atoms of oxygen per mole of fuel, which can be considered negligible. This oxygen requirement is likely also to be an overestimate since the equilibrium oxygen to metal ratio will be lower than the average value of the fuel. The. oxygen balance calculations adopting all these assumptions are detailed in table 3. The oxide forming fission products account for a fraction of the original oxygen and the balance, compared with the concentration of the matrix elements, provides the final oxygen to metal ratio. An alternative value is included for pin 2 allowing for clad corrosion. 4.2. Oxygen and temperature distribution The previous section was concerned with the overall change in oxygen-to-metal ratio of the fuel during
265
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsII
2.3 din
02 \ \ \
kcds.mole-' 04
\
\
03
\
2-i,_
\ IO 2.'
O’-
Oh
io-
21o-
L
1.9! j-
I
I
1
1
.2
.3
Fuel
radius
(cm1
Fig. 15. Oxygen potential versus fuel radius for each pin at end of life. 19 O-
I
I
.l
.2
Fuel
radius
I -3
Icml
Fig. 14.0/M ratio versus fuel radius for each pin at end of life.
irradiation. It is now necessary to consider the mutually interacting effects of temperature gradient and oxygen redistribution in that gradient. The redistribution of oxygen in an operating fast reactor has been discussed theoretically in a number of papers, discussed in the review by Bober and Schumacher [2]. These various treatments consider the oxygen distribution as the resultant of the effects of gaseous and solid state transport. This is considered to be represented by the equation: Inx=Q*IRT+c,
(1)
where x is the deviation from stoichiometry, R and T are the gas constant and absolute temperature respectively, c is a constant and Q* is the heat of transport for the oxygen transport process. A limited number of experimental observations of oxygen re-distribution
have been made on actual fuels [8,9] and in laboratory experiments on fuel materials subjected to temperature gradients [ 1O-l 21. These observations have given support over a limited temperature range to the theoretically derived expression for oxygen redistribution. Values of between 5 and 40 kcal/mole for the heat of transport Q* have been found under varying conditions of initial oxygen-to-metal ratio of the fuel. These values have then been used in the fuel chemistry computer modelling programme TPROF [ 131 to calculate the oxygen redistribution and oxygen potential within the fuel pin. This programme models fuel behaviour, it is used here to compute a temperature profile for the fuel and then to distribute the oxygen according to eq. (1). The thermal conductivity is then modified to accord with the calculated distribution; the programme repeats these calculations until satisfactory convergence is obtained. At each time step during the irradiation, the mean oxygen to metal ratio is calculated as outlined in section 4.1, and the temperature and oxygen profile determined. The results of these calculations are given in figs.
F. T. Ewart et al. / Composition
266
changes in irradiated
1000 Temperature
Fig. 16. Ellingham diagrams of
U, Pu
oxide fue! materialsII
1500
2000
(Cl
oxide [ 14,151, some fission product and clad component systems [161.
lo-15 where graphs are drawn of temperature, oxygen to metal ratios and oxygen potentials as a function of fuel radius for the start of life (figs. 10-12) and the end of life (figs. 13-l 5) conditions. The values of Q* selected for the computation of each oxygen-to-metal ratio profile are shown on each curve. The relationship between oxygen-to-metal ratio and oxygen potential used for these calculations is given in fig. 16. The values used are interpolated between those of Markin and McIver [ 141 and Javed [ 1S].
5. Discussion
The oxygen content of the fuel and its redistribution will result in differing conditions of oxygen potential which will control the local chemistry of the fuel and will be reflected in the type and composition of the segregate fission product phases formed. Since is has not been possible to measure local oxygen potentials in the fuel directly, a correlation has been attempted between the theoretical treatments described
and the observations metal inclusions.
of the composition
of the noble
5.1. Experimerztai indications of oxygen migration The oxygen redistribution can be illustrated by the equilibrium states of the oxidation reactions of the intermediate oxides. The two equilibria Mo+O*+ MOO, and Tc + 0, * Tc02 are particularly convenient in that the metals form segregate phases with the noble metals, and the oxides disperse into the matrix. The state of the equilibrium can, therefore, be readily detected with the electron probe micro-analyser. An approach to the quantitative interpretation of such data has been made to this problem previously by Johnson et al. [17]; such an interpretation is not attempted here, principally because of instrumental limitations. In pin 1 the MO and Tc in the metallic inclusions are present in ratios similar to their fission yields; the only indicator of local oxygen potential is the PuPd, phase formed in the fuel rim. In pin 2 there is clearly adequate oxygen available
261
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsII to oxidise the molybdenum
and a significant graction of the technetium also, this is evident from the ‘white’ inclusion composition. The existence of the molybdenum oxide phase in the insulator pellet suggests that the molybdenum has been oxidised to Mo03, a volatile oxide, which has distilled into the colder regions of the fuel. The 0, t 2Mo02 - 2Mo0, reaction is shown in fig. 16 and considering the extent of extrapolation of the thermodynamic data and the oxygen redistribution model to the O/M and temperature involved in this pin, it is not unreasonable to expect this reaction to occur. The data obtained here from pin 3 are in general agreement with the described models, that is, that an increasing fraction of the molybdenum is oxidised as the radius decreases; in pin 2 the system is much more oxidising, all the molybdenum has been oxidised and a similar gradient is observed in the technetium concentration. The observations from pin 4 are anomalous in that there is no detectable molybdenum concentration gradient, which is surprising when the evidence from ceramography was that the two types of pellet had equilibrated. There is, however, sufficient oxygen available in this pin to oxidise all the molybdenum. Some of the molybdenum has been retained in the metallic state, suggesting perhaps that some caesium oxide equilibrium acts as an oxygen sink before all the molybdenum is removed from the noble metal alloy. 5.2. Fission product segregation A large number of different fission product segregate phases have now been reported in the literature and have been reviewed by Kleykamp [ 181. In very few of the published works has it been possible to relate the existence or composition of the phases to the oxygen-to-metal ratio of the fuel or the irradiation conditions. The significance of this series of irradiations is that both the fuel and the irradiation conditions are well defined thus providing an opportunity to indicate the conditions of formation of a number of inclusion types. Table 4 lists the inclusions found and the estimated local oxygen potential deduced from fig. 16. The PuPd, type of inclusion found in pin 1 has been previously reported by Bramman [ 19 ] and Kleykamp [ 181 in sodium-logged failed fuel pins. These provide a relatively reducing environment which is consistent
Table 4 Fission product inclusions Pin no.
Inclusions
Temp (OC)
1
UtsPurrPd74 MoaTTctoRwRhrtPds P_ddUPu
1150 1700 700
Estimated oxygen potential (kcal/mole) -132 -125 -130
1950 T~~Rtre~Rh~~Pd~ 1950 PdwTeaaU4 1950 cs (01) 650 Tc MO (O?)* 650 Fe Cr Cs MO 6.50 FeCr Ni U Pu Cs Ba 650 pd Te Cs (U, Pu?) 650 PdeTesU *Volatile species originally in matrix
+ 20 + 20 + 20 -100 -100 -100 -100 -100
FeCr Ni Cs FeM &&Tc&R_hPd
900 1100
- 95 - 95
Pd74UaSnsTet4Cse Pd7eUsSnsTearCsr MO - Tc Ru - Rh Pd
1750 1000 general
2
3
4
general - 20 - 65 >- 80
Where analyses are not given, the principal constituents underlined.
are
with the oxygen potential calculated by Holleck [20] to be necessary for the formation of the URu3 , URh3 phases. The oxygen potential calculated was -125 kcal/mole, which shows some agreement with the value derived here (-130 kcal/mole) for the rim of the pin 1 fuel. The complex palladium-tellurium-caesium phases found in pins 2 and 4 are more unusual in that there is only one published reference to these phases (Huber and Kleykamp [21]). These workers found Pd, Te, Sn alloys in the columnar grains of 15% mixed oxide at an initial O/M of 1.98 which had reached 6.5% burnup. This fuel, which contained 11.5% 235U would have its chemistry controlled by the MO/MOO, equilibrium. The observation is in contrast to the present one where Pd, Te phases occur in the fuels which have exceeded this control. 5.3. Fuel-cladding interactions Although this experiment consisted of single pin irradiations, the results relate well to the experience
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materialsII
268
Table 5 Clad corrosion products ~__ ~~__ Pin no.
Inclusions
Oxygen potential (kcal/mole)
1 2 3 a) b) 4
CS 9 Ba Fe Cr Ni QFeCrG FePd Cs -_ Ba Te Nd MO
-120 - 90 ~ 90 - 90 - 90
Principal constituents
has been observed also by Kleykamp [ 181, no cladding elements were found in the fuel matrix remote from the clad. The Argonne work [25], for example, reports up to 13% iron in the noble metal inclusions and Crouthamel [ 171 has reported large numbers of iron rich inclusions deep in the fuel. The German work, however [18,21], reports Fe, Pd and Fe, Ni, Pd alloys near the fuel surface and in one case [ 181 reports a clad interaction similar to that found in pin 2, but occurring in a fuel pin of starting O/M 1.98 f 0.01.
are underlined.
5.4. Heavy metal concentrations found in large-scale irradiation experiments. The modes of attack, intergranular and broad front have been reported previously 121-231 in considerable detail. The results from these experiments support the belief that clad attack can be minimised by low oxygen potential conditions at the clad and may have some bearing on defining the conditions that give rise to intergranular or broad front attack. Table 5 summarises the analyses of the reaction products in the fuel-clad gaps. The predominant species found in these pins was caesium. In pins 2 and 3, where the final oxygen to metal ratio had exceeded the control offered by the MO/MOO, equilibrium, barium was also present. The observations on pin 1 suggest that caesium is compatible with the clad at a local oxygen potential of - 130 kcal/mole. The difference in the rim oxygen potentials of the other pins (-~90 to -9.5 kcal/mole) are small and do not account for the differences in corrosion mechanisms. However, the enrichment of chromium in the reaction product in pin 2 may be significant, in that interaction of caesium oxide and chromium could remove the protective oxide layer from the steel and would allow a ‘broad front’ attack. There is evidence [24] that the Te : Cs ratio may change the oxygen potential at which a particular mode of clad attack operates. The differences in corrosion phenomena found here in fuels pins having apparently similar oxygen potentials at the clad may be accounted for by the widely differing oxygen potentials in the body of the fuel which could control the relative rates of release of tellurium and caesium to the fuel-clad gap. A significant difference between these irradiations and those reported by other workers [20,22,25] is in the fission product-clad component phases. With the exception of the Fe, Pd phase found in pin 3, which
Some consideration must be given to the composition of the matrix in Pu-enriched fuels which are irradiated in thermal fluxes to high burn-up. The assumption is made, during discussions of fast reactor fuel, that the chemistry of the matrix is controlled by the valency state of the uranium (O/M > 2) or plutonium (O/M < 2). In thermal irradiations, however, considerable Pu burn-out occurs and the lanthanide concentration in the matrix becomes a significant (fig. 17) fraction of the total presenting a differing chemical situation from that in a fast reactor. This effect is demonstrated in fig. 17. For an MTR irradiation, the majority of fissions are in plutonium and plutonium replenishment by neutron capture is relatively unimportant. This is compared with a Dounreay Fast Reactor (DFR) irradiation where fissions are in both uranium and plutonium and with a Prototype Fast Reactor (PFR) irradiation where fissions are in plutonium and there is a significant Pu replenishment rate. The chemistry of the resulting system is thus complex and is controlled by the solid solution of U, Pu and lanthanides. There are very few thermodynamic data available for solutions of lanthanides in Urania, plutonia or mixed systems. The observations of Markin and Crouch [26] show that the oxygen potential of ceria-Urania solution is higher than that of equivalent plutonia-Urania solutions in the hypostoichiometric regions; the observations of Stadlbauer [27] in the hyperstoichiometric region of lanthania-Urania are in agreement with this trend. At this stage, the available data are inadequate to allow a quantitative assessment of the effects of Pu burn-out. There is, however, a reasonable indication that the oxygen potential of the oxide matrix will be,
F. T. Ewart et al. / Composition changes in irradiatedoxide fuel materialsIf Information corrosion
DFR 15%Pu 79 % 235u
269
on the cause and mechanism
is difficult
to achieve
of clad
on the basis of single
pin experiments. It can be concluded, however, that an oxygen potential at the clad of less than -130 kcal/mole is likely to inhibit clad corrosion. It is also apparent that the distinction between intergranular and broad front corrosion, both of which occur at -90 to -95 kcal/mole, is not simply attributable to the oxygen potential at the surface of the fuel but is a function of the chemical species present in the fuel/ clad gap. These will depend on the transport mechanisms available and the oxygen potential elsewhere in the fuel.
Lonthonsde
Acknowledgements MTR 15% Pu
Fig. 17. Concentration of piutonium and lanthanide elements as a function of burn-up.
in the case of fuel irradiated in thermal fluxes, significantly more oxidising than the equivalent fast reactor irradiation because of the higher rate of plutonium burn-out.
Acknowledgement must be made to the considerable skills in sample preparation shown by the Process Workers of Harwell without whom this work would not be possible. The experimental assistance with the EPMA of Mr. L. Murphy is also acknowledged. Dr. F. Caligara of Euratom, Karlsruhe is thanked for allowing the use of his computer code ‘TPROF’ for fuel performance modelhng.
References 6. Conclusions The general trend of observations made on the chemical form of the fission products is in agreement with a model of increasing oxygen-to-metal ratio with increasing burn-up. The variations in composition of the noble metal inclusions with fuel radius conform in general with theories of oxygen redistribution and the particular observations of the PuPd, phase are in good agreement with the model of oxygen redistribution used. In order to obtain quantitative data relating to oxygen redistribution and test the validity of the model, there is a need for improved experimental techniques of oxygen potential measurements. Development of techniques for measurement is of ~portance, but a deal of information would be available if the conditions of formation of the various fission product segregate phases were known.
[l] J.H. Davies and F.T. Ewart, J. Nucl. Mat. 41 (1971) 143. [ 21 M. Bober and G. Schumacher, Adv. in Nucl. Sci. and Tech. 7 (1973) 21. [3] J. Bramman and H.J. Powell, J. Brit. Nucl. En. Sot. 14 (1975) 63. [4] W. Batey and K.Q. Bagley, J. Brit. Nucl. En. Sot. 13 (1974) 49. [ 51 J. Adam and F.T. Ewart, BNES Symposium on Post-irradiation Examination Techniques, Reading f1972). [61 D.M. Poole, UKAEA Report AERE-2469 (1971). [ 71 F.T. Ewart and H. Kaneko, UKAEA Report AERE R7191(1972). [8] C.E. Johnson et al., Reactor Technology 15 (1972-73) 303. [9] E.G. Aitken, S.K. Evans, M.G. Adamson and J.H. Davies, IAEA Symposium on Thermodynamics of Nuclear Materials, Vienna (1974). [lo] A.E. Aitken et al., J. Nucl. Mat. 30 (1969) 57. [ I1 ] M.G. Adamson and R.F.A. Carney, UKAEA Report, AERE R-6831 (1972). [ 121 M.G. Adamson and R.F.A. Carney, UKAEA Report, AERE R-6830 (1972).
270
F. T. Ewart et al. / Composition changes in irradiated oxide fuel materials I1
[ 131 F. Caligara, Euratom, Karlsruhe, private communication
[ 141 [ 151 [ 161
[ 171 [ 181
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