Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

132

Journal

of Nuclear

Materials 166 (1989) 132- 159 North-Holland. Amsterdam

SAFETY ASPECTS OF FUEL BEHAVIOUR DURING FAULTS AND ACCIDENTS IN PRESSURISED WATER REACTORS AND IN LIQUID SODIUM COOLED FAST BREEDER REACTORS J.H. GITIUS

‘, J.R. MATTHEWS

* and P.E. POTTER

3

’ UKAEA Director, Communication and Information, Charles II Street, London S WI Y 4QP, Unrted Kingdom ’ Theoretical Physics Division. Harwell Laboratory, Didcoi, Oxfordslure OXI I ORA, United Kingdom -’ Chemistry Division, Harwell Laboratory Didcot, Oxfordshire OX11 ORA, United Kmgdom

The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO, based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified.

1. Introduction Over the past twenty five years in the countries of the European Community, nuclear power has become a major, and in some countries the dominant, source of electricity. During that period there have been no significant accidents or releases of radioactivity from power reactors, although we have seen the major incidents outside the Community. This good safety record is partly based on sound reactor design and partly on responsible reactor operation, but some credit should also be given to the large effort given to understanding the safety characteristics of the reactors and their fuel, which has been obtained by the efforts of the Joint Research Centre and the national nuclear research institutions, often in collaboration with partners outside the Community. In this paper we wish to review briefly the present state of our knowledge of oxide fuels with respect to the safe operation of reactors. Nearly all the current European reactors use fuels based on the oxides of uranium and plutonium. We will divide the paper into a part concerned with water reactor fuels based on UO, and a part concerned with fast reactor fuel based on the mixed oxide of uranium and plutonium. We will not discuss specifically the behaviour of UO, in gas cooled reactors. Although they share many characteristics with

light water reactor fuels, there are some special aspects which we do not have space to deal with. The main theme of the paper is the importance of the role of the fission products and the chemical state of the fuel on all aspects of fuel behaviour. We will discuss the factors controlling the integrity of the fuel during normal operation, the consequences of fuel failure and, the response of the fuel to reactor faults and structural failures and finally the consequences of a large accident leading to substantial damage to the reactor core. The analysis of faults and accidents places considerable demands on our understanding of chemistry and properties of fuels. Thus an essential step in the prediction of the release of radionuclides in a fault or accident is the possession of a detailed knowledge of the fuel temperatures and the chemical constitution at all stages of the development of the incident. Before going on to describe these specific points it is worth looking at a common feature which affects consequence analysis for accidents in all reactor types. In order to assess the consequences of faults and accidents an evaluation of the likely releases of the potentially hazardous radio nuclides has to be made. The chemical state of these radionuclides is a dominant factor in determining their fate. A list of the radionuclides which have to be considered in degraded core accidents is shown in table 1, which is taken from the Reactor

J. H. Gittus et al. / Safety aspects Table 1 Radionuclides which have been considered in consequence analysis for degraded core accidents in light water reactors (ref.

PI) Element Krypton Rubidium Strontium Yttrium Zirconium Niobium Molybdenum Technetium Ruthenium Rhodium Tellurium Antimony Iodine Xenon Caesium Barium Lanthanum Cerium Praseodymium Neodymium

Neptunium Plutonium Americium Curium

Radionuclide SSK,

85mG

87~.

88~~

s6Rb 89Sr 3“Sr 391Sr WY, 9’Y 95Zr, 97Zr 95Nb 99Mo WmTc VJXRu 105, ‘“Ru IO& ‘2,Te, %T,, ‘29Te, 12TSb ‘29Sb 2311

;3q

1331

1341

,33i,

‘3iXe

’

‘34cs,

=%,

137cs

,3,“lTe,

132Te

fuel rod during operation has attracted much attention. Thus, there has been a considerable amount of research on irradiated fuel and on material in which irradiation is simulated by the addition of the fission product elements as the non-radioactive nuclides. We shall consider in this section: (i) The chemical constitution of the fuel during operation at temperatures corresponding to quite low ratings, together with the appropriate information on thermodynamic parameters and phase equilibria. (ii) The mechanical and structural response of the fuel to operation, particularly during power changes. (iii) The changes in constitution on failure of fuel rods in fault and accident conditions. 2.1. The chemical constitution of fuel during operation

1351 ’

The fuel for a light water reactor is essentially stoichiometric uranium dioxide. The conditions under which the fuel operates can result in fuel centre temperatures as low as ca. 1200 o C. UO, possesses the fluorite structure and has the ability to accommodate readily interstitial oxygen anions to form hyperstoichiometric and also to lose oxygen anions resulting in UO,+, hypostoichiometric UO,_,. Much attention has been given to the crystal chemistry of UO,,,, [2-41. Parts of the phase diagram of the uranium-oxygen system are shown in figs. 1 and 2; the diagram is divided into two regions, one with the uranium : oxygen ratio > 2.0 and

lmBa ‘@La ‘41 Ce, ‘43 Ce, ‘*Ce 143Pr

14’Nd 239Np 238

Pu, *39Pu, *4oPu, 24’Pu 241Am 242Cm, 2”Cm

Safety Study of the US Nuclear sion [l].

133

offuelbehaviour

Regulatory

Commis-

2. Water reactor fuel behaviow In the first part of the paper we will discuss those aspects of the fuel behaviour that are important for the safety of water reactor fuel. The fuel for water reactors consists of pellets of Urania (UO,), enriched in 235U, clad in Zircaloy; in the future fuel containing plutonia will be utilised in such reactors as a solution of up to 6% plutonia (PuO~_~) in Urania. The relatively low coolant temperature and moderate linear ratings of around 35 kW/m result in a low fuel centre temperature, generally below 1200 o C, the consequence of this is that most of the fission products are retained within the fuel matrix during operation, although a small fraction of some of the more volatile species are released to the fuel-clad gap. Fission of the actinide elements produces over thirty new elements and the way in which these fission product elements are distributed within the fuel and

l-d

0

Fig.

1. Partial

5

phase

diagram of the uo,_,.

U-O

system,

region

J. H. Gitius et al. / Safety aspects of fuel hehaoiour

134

^“‘t 20

‘;,

y2+ay

,

21

22 o/u

I

I

Fig. 2. Partial phase diagram of the U-O uo2.x.

system, region

the other with the ratio < 2.0 [5]. The positions of the upper phase boundary of UO,-, are shown in fig. 1, and because the energy required to form a vacancy in the oxygen lattice of UO, is much greater than that involved in the formation of interstitial oxygen atoms, the value of x in UO,_, at the lower phase boundary is very close to zero up to ca. 1500 K, but then increases to about 0.33 at the monotectic temperature, 2700 K (fig. 1). The phase relations for UO,+, are shown in fig. 2. There is a marked change in the thermodynamic potential of oxygen in the Urania phase at an oxygen : uranium ratio of ca. 2.00 (fig. 3); these data are cited in a recent review on phase equilibria and thermodynamics of oxide nuclear fuels [6]. The thermodynamic oxygen potential is of great significance in determining the chemical constitution of irradiated fuel. The diffusion rates of many different atoms and ions within the fluorite lattice are markedly influenced by the ratio of oxygen anions : metal cations of the lattice [7]. Because of the significance of these effects of thermodynamic oxygen potential and composition of the oxide, in addition to experimental determinations, there has been much activity in the devel-

opment of models which reproduce the experimentally determined values of thermodynamic oxygen potential (Go>) for different temperatures and composition. Such models are required for the extrapolation of data outside the ranges of measurements. A brief description of the earlier models was given by Potter and Rand [8]. The defect models [g-13] were based on the formulation of Anderson [14] for the construction of a grand partition function for defect solids; later work has mainly focussed attention on ‘extended’ or non-random defects, electronic defects, and the inclusion of substitutional atoms. The latter aspects have been considered by those authors [15-211 who avoid the use of a partition function by assuming a particular defect equilibrium within the lattice and making use of experimental data to determine the requisite equilibrium constants. The approach has the merit of calculational simplicity and has been able to represent the oxygen potentials of a number of oxide systems, particularly those containing UO,, with significant success. Before discussion of the recent studies on these models, complementary work on the computer simulations of the ionic and electronic defects which gives good indications of their energies must be mentioned; Jackson et al. [22] have reviewed the studies on the calculation of defect parameters in UO,. There are also discussions on the methods of calculation of defect energies [23] and entropies [24] and some further developments of statistical mechanical models [25,26]. A particularly noteworthy contribution on the application of specific defect equilibria is that of Hyland [27]. The chosen defects are U3+, U5+, oxygen interstitial atoms and vacancies. The defect equilibria are written in a rigorous way and differentiate between oxygen atoms on lattice and interstitial sites and between lattice and interstitial anion vacancies. A major advance is a description of the defect equilibria for the intrinsic region, close to the stoichiometric composition, in which the equilibrium concentrations of U3+ and U ‘+ cations together with those of the vacancies and interstitial oxygen anions exceed those due to any non-stoichiometry in the lattice. It was found essential to re-examine the published experimental information in order to obtain a consistent interpretation of the data for co2 of UO,,,. Hyland analysed the experimental data which extended either side of the stoichiometric composition [28,29] (fig. 3). The analysis of Hyland gave: Go1 = 786.6 + 167.4t + (210.1 + 525t2)“2 kJ mall’ where

0,.

t = T(K)/lOOO,

for

the

variation

of

oxygen

J.H. Gittus ei al. / Safety aspects

xm 15

-10

-5

0

5

10

15

20

25

30

35

40x10-3

Fig. 3. The variation of oxygen potential (co*) with composition for Urania. Data at 1273 K from Baranov and Godin [29]. The curve was obtained from the model of Hyland (27). x, is the measured deviation from stoichiometry. Go, is in kcal/mol O,, 1 cal = 4.184 J.

temperature (1.273 < 1.950). This equaused by Ohse et al. [30] to extrapolate Go1 up to the melting point (3120 K) of UO,,,. The predictions of this model are in excellent agreement with those calculated by an earlier model of Blackbum [15]. Another approach using defect equilibria has also been described by Ohse et al. [30,31] in which the hypostoichiometric phase was described in terms of a mixture of U(I) and UO, and the hyperstoichiometric phase in terms of a mixture of UO, and UO,. Lindemer and Besmann [32] have represented all the measurements which have been made of thermodynamic oxygen potential for hypo- and hyperstoichiometric Urania. Representations of the nonstoichiometric regions of Urania are very convenient to use in the calculations of multicomponent equilibria involving the Urania phase. Under irradiation, the fuel will be in a temperature gradient; the magnitude of the gradient will depend on the rate of fission and the conductivity across the gap between the fuel surface and the cladding wall. The temperature gradient would result in transport of material and, as already mentioned, for fuel with relatively low ratings and thus with low centre temperatures (say < 1200 o C) the irradiated fuel will be a solid solution of the fission product ions and elements in the Urania lattice; little nucleation of separate phases would occur at such temperatures. The average valency of the fission product elements and cations will be less than four and some oxidation of the U4+ cations to U5+ potential

with

tion has been

offuel behaviour

135

cations will occur resulting in an increase in the thermodynamic oxygen potential of the fuel matrix with increasing burn-up of the actinide atoms. Such an increase in oxygen potential would only occur providing that there be no transfer from the fuel matrix to the Zircaloy cladding. In order to allow some quantitative assessments to be made of the increase in oxygen potential with burn-up there have been several determinations of the relationship between composition and oxygen potential of the solid solutions of the lanthanide oxides in Urania and of the Urania-plutonia system. Lindemer and Brynestad [33] have recently reviewed and presented a representation of the variation of thermodynamic oxygen potential with composition and temperature for the systems U,_,Ce,O, f * and with Ln = Y, La, Nd and Gd. In the Ui-,Ln,Gz,,; solid solution with ceria, the cations can have valencies of 3 and 4, whereas in the latter group the Ln cations possess a unique valency of 3. The incorporation of the trivalent lanthanide cations into the Urania lattice at a given oxygen anion: metal cation ratio results in an increase in co2 of the solution with respect to Urania of the same oxygen anion : uranium cation ratio. Similar behaviour occurs for the U, _ICe,O,p, solid solution. The Urania-plutonia solid solution is analogous to the Urania-ceria solution; there is an appreciable range of composition of fluorite structured solid solution of Urania and plutonia, in which the U4+ cations can be oxidized and the Pu4+ cations reduced [34]. At temperatures above 1500 K, the U4+ cations can also be reduced. Considerable attention has been given to the measurement of the variation of thermodynamic oxygen potential of the solid solution with composition and temperature; measurements have been essentially confined to solutions with Pu : U ratios of technological significance to their application as fuels for fast breeder reactors (see section 3). Most of these measurements have been fitted to a simple model by Lindemer and Besmann [35,36], but there are some discrepancies between predictions of this model and other models [16,37,38]. Whilst the models can be used to predict the oxygen’ potentials of solid solutions with plutonia concentrations appropriate to the utilisation of plutonium in light water reactors, we still require more experimental information on the variation of Go2 with composition of the solid solutions at temperatures below 1200 ‘C. There is no information on solid solutions with plutonia concentrations less than 11%. Woodley [39] measured the oxygen potential of Urania-plutonia solutions (25 mol% plutonia) to which oxides of Y, Zr, Ce, Pr and Nd were added to simulate bum-ups of the actinide atoms of up to 10% in the

136

J.H.

G‘mus et al. / Safely aspects oJJuel behavlour

temperature range 900-1100 o C using a thermogravimetric technique. For hypostoichiometric solid solutions of a given oxygen anion : metal cation ratio, the oxygen potential increases linearly with simulated burnup. As the oxygen potential increases within the fuel matrix even in relatively low temperature gradients with fuel centre temperatures of 1200-1400 o C, some redistribution of oxygen could occur by both gaseous and solid state mechanisms [40-431. Rand and Roberts [40] in a classic paper on the application of chemical thermodynamics to the interpretation of the changes in chemical constitution of oxide fuel during irradiation suggested that a possible mechanism for the transfer of oxygen within an oxide nuclear fuel was through the gas phase with mixtures of CO/CO, or H,/H,O. The presence of the gases would be maintained by thermal convection through interconnected porosity in the fuel, and the fuel, at every temperature, would be in equilibrium with the gaseous mixture. Aitken [41] developed a more general approach in which solid state transport of oxygen is also considered, and indeed as the oxygen anion: metal cation of a solution of Urania-plutonia decreases this latter mechanism becomes more significant. The gas phase mechanism will be predominant in hyperstoichiometric oxide fuel. Such a mechanism predicts the observed direction of oxygen movement towards higher temperatures. The transport of the actinide and fission product elements can also occur in fuel pins due to such temperature gradients but we shall not be concerned with such phenomena here as the temperature gradients in modern operating LWR fuel would not be high enough to cause significant transport of species other than that of the most volatile fission product elements. Adamson et al. [44] have measured the radial distribution of oxygen in irradiated fuel rods of Urania clad in Zircaloy-2. For lightly irradiated fuel of nominally stoichiometric composition and with an 0: U ratio of 2.033 irradiated at the high rating of 525 W/cm (estimated centre temperature 1820 o C), oxygen migrated to the centre of the fuel but for the higher stoichiometry some transport of the oxygen to and reaction with the Zircaloy cladding had occurred. Examination of nominally stoichiometric Urania irradiated up to 18900 and 11400 MWd/t at ratings of 230 W/cm and 394 W/cm respectively indicated that the oxygen potential in the centre of the fuel had decreased for both conditions; the oxygen had probably reacted with the cladding and with caesium. The technique which Adamson et al. [44] used to measure oxygen potential employed an EMF solid state cell with a thoria-yttria electrolyte and a Ni-NiO reference

electrode. This technique has also been used to measure the oxygen potential of irradiated fast reactor oxide fuel [45] further. An important aspect of fuel performance is the maintenance of the integrity of the cladding and there have been considerable efforts to understand the mechanisms of mechanical and chemical interactions which can result in the loss of such integrity. In considering any chemical reactions between fuel and cladding, a knowledge of the gap inventory of fission product elements is required; and of the mechanisms leading to their presence. In fuel, under normal operating conditions, the presence of fission product elements in the gap is largely a result of fragments from fission and the inventories are very low, probably less than 1% of the total amount of fission product elements. At higher temperatures, during normal operation. and during transient ingresses of energy and in subsequent heating in accident conditions the predominant mechanism for the loss of fission product elements from the fuel matrix is that of diffusion. Applications of the technique described to examine the changes in oxygen potential which occur during irradiation of fuel should be considered particularly for fuel irradiated at appropriate ratings. We have indicated that the oxygen potential of a fuel matrix will increase with the extent of irradiation, although the precise quantification of this change is difficult to predict because of diffusion of the oxygen to the cladding. As the rating of the fuel increases, the rates of diffusion of the fission product elements will increase. The fission product rare gases Kr and Xe have been considered chemically inert, but recently MacInnes and Winter [46] have suggested that if these fission product species reside in defect clusters they will bind oxygen and cause a depletion in the concentration of interstitial oxygen atoms in the oxide lattice. and thus reduce the oxygen potential of the fuel. It is important to have an understanding of the relationship between oxygen potential and the various parameters of fuel, for example rating and burn-up of the fissile atoms, because the oxygen potential can so markedly influence the transport properties of the components of the oxide lattice The appropriate processes which determine the release and behaviour of the fission product elements and ions are diffusion in the grains of Urania, in the case of the inert gases some intragranular diffusion of bubbles occurs to the faces and then to grain edges where the gas can be released through interconnected porosity formed from the linking of fission gas bubbles and porosity from the fabrication process. Olander [47] has

J. H. Giltus et al. / Safety aspects of fuel behauiour

recently described a model which considers both grain boundary and volume diffusion in fine grained polycrystalline bodies. This model considers two release paths for the fission product atoms and ions; the first is that of volume diffusion and the second is lattice diffusion from the grain interior to the grain boundaries followed by migration in the grain boundaries to a free surface. The properties which control the release of any fission product atom or ion are its diffusion coefficient in the lattice of Urania and the product of its diffusivity in the grain boundary and the distribution coefficient of the atom or ion between the lattice and grain boundary. The relative importance of the two pathways depends on these transport properties and on the grain size. Models have been developed to allow predictions of release of the inert gases at different power ratings of the fuel and for conditions of power transients. The distribution coefficients for the various atoms and ions will determine the chemical speciation of the fission product elements on the grain boundaries. In addition to Kr and Xe, the fission product elements Cs, I and Te are readily lost from irradiated fuel heated at temperatures over 1500” C. Thus, if these individual fission products collect on the grain boundaries it seems appropriate to consider their chemical constitution within the grain boundaries and the influence of this speciation on their subsequent behaviour and release from the grain boundaries to the fuel-clad gap and to the water and steam atmospheres if the cladding were to fail. For the release from the grains by volume diffusion directly into the gap, the chemical speciation would not change from that in the grain boundaries, but when released directly into the water or steam atmosphere the fission products may be in elemental rather than combined form. There is evidence from the measurements of release of fission product elements from irradiated uranium dioxide that in addition to the rare gases Kr and Xe, the elements Cs, I and Te are also readily lost from the matrix of overheated fuel [48]. The three latter elements will appear on the grain boundaries of the fuel; the fission yields are in the proportions 18 : 2.3 : 1. Clearly Rb, Br, and Se would also be present, in that they have similar properties to the Cs, I and Te but because their fission yields are much lower, they have been neglected in our considerations. The extent of release of these elements from the grain boundaries in the fuel-clad gap would depend on their chemical constitution in these locations. In order to predict and understand such behaviour, a knowledge of the phase relationships of Cs with UO,, I and Te is required. We have already noted

137

that changes in the thermodynamic oxygen potential of the fuel matrix will occur during irradiation and it is therefore essential to understand the influence that such changes in oxygen potential have on the phase relationships. In order to model the constitutional changes which can occur during irradiation and also to interpret the post-irradiation examination of irradiated fuel, of failed fuel and of debris from experiments which attempt to simulate degraded cores, it is essential to have a detailed knowledge of the appropriate phase equilibria. We require a sound collection of thermodynamic data for the condensed phases, solutions, and gaseous species. Indeed much of this required data is available but there is always a requirement to fill important gaps in such data. We shall require the thermodynamic quantities for all the condensed phases or compounds and for all the gaseous species which are likely to be encountered at the thermodynamic potentials of oxygen appropriate to irradiated fuel. With these thermodynamic quantities we shall obtain for a given overall composition of the elements an assemblage of all the possible condensed phases together with the composition of the gaseous phase which gives the minimum total Gibbs energy of the system. In order to calculate this minimum the program SOLGASMIX developed by Eriksson [49] has been widely used. We shall illustrate the potential of this method by describing the changes in constitution within the fuel-clad gap of a PWR fuel rod when the oxygen potential is changed. For this assessment of the chemical constitution of Cs, I and Te with Urania, we need information about the thermodynamic quantities of the condensed phases and species in the gas phase which could be present. The data for all these condensed phases and gaseous species are taken from a recent assessment [50], and other published data [51,52]. The model used for the uo 2+x phase was that developed by Lindemer and Besmann [32]. Aspects of the chemical constitution of LWR fuel rods have been considered by Besmann and Lindemer [53] and more recently by Burns et al [54] and Potter [55]. For the gap, we have considered a release of Cs, I and Te to the gap of 1% of the total fission product inventory equivalent to an average burn-up of the uranium atoms of 2.9%; the fuel rods are filled with helium gas. In fig. 4, we give some results of a calculation of the speciation in the gap at 650 K. We have successively added increments of lop5 mol 0 atoms to the system, which increases the oxygen potential of the system, and

138

J. H. Gitttu et al. / Safety aspects of fuel behauiour Temperature

650K

10 r

1% inventory

Condensed

In gap

to note is that the potential of iodine increases with increasing oxygen potential (> - 700 kJ/mol) but the pressure of I(g) is always considerably lower than that of CsI(g). It is unlikely that any stress corrosion cracking of the Zircaloy cladding could be due to iodine formed by these thermal reactions. Under reactor operating conditions, we can expect concentrations of the gaseous species to be greatly changed from those at thermal equilibrium because the large continuous input of radiation energy, largely in the form of fission fragment radiation, tend to dissociate molecules into neutral atoms or into ions. This has been recognised by Konashi et al. [56,57] and some expected features of the processes, such as the dependence of the presence of iodine atoms on the square root of the dose-rate, in a mechanism containing recombination as the ultimate fate of transient species. have been suggested by Kanno et al. [58]. An important aspect of the behaviour is that we are dealing with a dilute mixture of other gases in helium [54]. Helium is well known for greatly sensitising the radiolytic decomposition of minor components of its mixtures. This occurs by way of charge and energy transfer from its highly energetic ion He+ and metastable atomic states 23S and 2’s. All of these species will take part in inelastic collisions with gas phase molecules, including CsI and I,. easily dissociating them. The combined effect of all the reactions which take place due to irradiation will cause a new dynamic equilibrium to be established quickly for which we expect the steady state concentrations of I atoms to be greater than in the absence of radiation.

Phases

-18

-26 1

900

500 KJ mot-’ O2

700 -60~

I

300

Fig. 4. The variation of the chemical constitution fuel-clad gap of a PWR fuel rod at 650 K.

of the

have calculated the change in the pressures of the gaseous species and which condensed phases are formed with increase in oxygen potential. An important point

No radlatlon Or Pressure CsI Temperature

-5 ; ;i ? -10 “9 E : -15

q

q

173 x 10m3 PA 713 K

-r

Pressure I for of Zlrcaloy

XC

;i: -20 .r ;

--

I

-25 k _ _ --__

x-30

t

--__

1:: -35 I

--__

--__

--__z I

--__

-LO 1 -5

-&

-3

Fig. 5. The thermal

--__

I

I

-2 0 -1 Log [pressure cs / pa )

1

I

equilibrium

--__ Lp -_ ~~

for the gas phase over CsI (solid).

2

139

J. H. Gittus et al. / Safety aspects of fuel behaviour

Neutral

species

‘\

Pressure I for SCC _ of Zircaloy

-

‘\ ‘\\

‘\

Pressure CsI Temperature -IL

1

q

1 73 x 10m3PA

q

713 K

,

-5

\

\

\ i 2’

\

‘\ ‘\ \

.bL_

-L

-3

‘\

\

I

-2 -1 0 Log (pressure cs /pa 1

1

2

Fig. 6. The effect of the radiation field on the pressures of the neutral gaseous species over CsI (solid).

Burns et al. [54] computed the influence of radiation on the equilibria in the gap using the FACSIMILE program [59] which formulates and solves numerically the necessary differential equations in species concentrations with respect to time when the reactions, rate constants, and starting concentrations are provided. For the example given, the mean gap temperature was 713 K, and the assumed pressure of CsI(g) was that in equilibrium with the solid at 650 K, the cladding surface temperature, and we have made wide variations in the partial pressure of Cs. Fig. 5 shows the variation of I

Neutral

and I, pressures with partial pressure of Cs in the absence of radiation. Fig. 6 shows how the computed partial pressures for reactants, except those of He and CsI which remain virtually constant, vary with the Cs pressure in the radiation field. The calculated pressures of iodine are considerably greater than those quoted by Konashi et al. [57] as the threshold for iodine to cause stress corrosion cracking of the Zircaloy clad. The effect of ionic reactions was also considered. The rate constants were estimated as typical of their types, which are charge transfer, dissociative positive

and ionlc species

Or

1.

‘.. I2

‘\

Pressure I for SCC of Zlrcaloy “\ ‘\

-u E

J -20 -

‘\

Pressure CsI = 1 73 x 10m3PA Temperature q 713 K

\. \

I I -5

\ 1

-L

1

i

I

1

,

I \

-3

-2

-1

0

1

2

Log (pressure

es/pa)

Fig. 7. The effect of the radiation field on the pressures of the neutral and ionic gaseous species over CsI (solid).

J. H. Citrus et al. / Sufety aspects of fuel behauiour

140

ion-molecule reactions, dissociative and non-dissociative electron capture and recombination with electrons or with negative ions. The final effect of all these reactions is shown in fig. 7 in which it is seen that the stronger tendency towards recombination in the conditions where ionic dissociation is considered, causes lower calculated concentrations of I, however the concentrations of I, will again be greater than without the radiation fields. In these studies, wall effects and the participation of a number of other potential species and reactions are neglected. and possibly Cs,Te and Species include Cs,, Cs,I, reactions include those with Xe. As well as displacing the thermodynamic equilibria in the gas phase, it must also be conceded that the effect of such high radiation dose rates may also disturb condensed phase equilibria ]601. In our discussions of the chemical constitution of the fuel-clad gap we have not discussed either the interaction between the fuel and cladding or additional volatile species. We have briefly mentioned the possible reaction of fuel with Zircaloy; the oxygen produced in fission can react with the cladding [44] in which it dissolves and could also form ZrO,. In order to form the phase Cs,ZrO,, the oxygen potential would have to be greater than that required to form ZrO, [61,62]. Although this phase has not been positively identified, evidence for the association of caesium with zirconia and mania in the fuel clad gap has been presented [63]. Although elemental molybdenum is not volatile, its oxides are. Molybdenum was considered to play an important role in the buffering of the oxygen potential of, for example, fast reactor fuel [64] in which the ratings would normally be considerably higher than those for LWR fuel. We shall return to this aspect of fission product chemistry later. For the lower rated fuel, if a small amount of MO were available to react with Cs and 0, the very volatile molecule Cs,MoO, could be formed [65] (vapour pressure at 1000 K, ca. lo-’ bar). We have now given an indication of the likely behaviour of Cs. I and Te within the fuel clad gap. If the fuel experiences power transients then additional amounts of fission products would be found in the gap and on the grain boundaries of the fuel. It is also possible that phases additional to those considered here could be formed; such phases would be characteristic of those which would be found in more highly rated fuel [66]. 2.2. Fuel performance

and failure

The trend in the development of water reactor fuel has been to go to lower operating ratings, hence to

lower fuel temperatures with consequently higher fission product retention and little fuel thermal restructuring. In these conditions thermal transport processes are inhibited and fission damage effects are relatively more important; leading to the fuel being in a state far from thermal equilibrium during steady operation. On entering a power transient, whether produced by an intentional operational power change or as a result of an abnormal event, substantial changes in both the chemical and microstructural state of the fuel occur, particularly in the central regions of the fuel where the temperatures are large enough for kinetic processes to be sufficiently rapid. During steady state operation one of the prime concerns is the stability of the fuel. Excessive densification in long fuel stacks can lead to reactivity effects, while radial contraction increases fuel temperatures that can in turn lead to further densification or alternatively result in clad collapse. Incorporation of some porosity in the fuel is desirable for high burn-up operation in order to accommodate fission product induced swelling and to cushion the effects of pellet-clad mechanical interaction (PCI) on power changes. Fine porosity is eliminated rapidly by interactions with fission fragments at even the lowest temperatures seen in the fuel (67,681. Pores less than 1 pm in diameter are removed completely after a fraction of 1% burn-up of fissile atoms. Pores between 1 and 10 pm in diameter are removed more slowly, while pores greater than 10 pm in diameter are essentially insensitive to fission fragment effects and require high temperatures (1 1400 o C) for their removal by thermal sintering. To take advantage of this, care has to be taken during fuel fabrication to avoid large volume fractions of the fine pores, but often large spherical pores are created deliberately by the simple means of the addition of some sacrificial material in the form of spheres (typically 50 pm in diameter) to the powder feed during fabrication. The form of the fuel porosity is also extremely important in determining fuel temperatures. The fuel thermal conductivity is not only sensitive to the volume fraction of pores but also the shape of the pores [69]. Spherical pores have the least effect, while flattened into connected porosity they can have five times the effect on the thermal conductivity for the same volume fraction. The thermal conductivity will thus change as sintering of fabrication porosity and swelling from the growth of fission gas bubbles takes place. These effects, together with the effect of soluble fission products, lead to a gradual increase in fuel temperature for a given fuel linear rating as burn-up accumulates. Soluble fission products degrade the thermal conductivity by increasing

J. H. Gittus et al. / Safety aspects

5480Cell

Parameter

Low

gap. Release of fission gas is thought to be a diffusive process and there has been considerable success in calculating release during steady operation using gas diffusion coefficients measured in reactor experiments by a variety of techniques [76,77]. Unfortunately these calculations were not found to be so successful when applied to the analysis of the behaviour of fuel rods subjected to mild power transients. One possibility suggested to explain the lack of agreement was that the fuel temperatures were being underestimated because of the effects on fuel thermal conductivity described above. However, the consensus among many workers in the field is that the effective gas atoms diffusion coefficient is a function of fuel burn-up and that the diffusion rates are approximately a factor of 50 higher at 4% fissile atom burn-up [77-791. The way this could occur is not yet fully understood but it might arise from direct effects of soluble fission products or more likely as a result of changes in the fuel stoichiometry during burnup. It is usually expected that the fuel oxygen to metal ratio will increase with burn-up. This result is based on an assessment of the chemical form of each fission product within the fuel at thermodynamic equilibrium. Recently published information indicates that PWR fuel has an increasing lattice parameter with burn-up, see fig. 8, which would indicate a decreasing oxygen to

(nm)

oxygen

--_

141

processes, fuel swelling occurs which can increase PC1 effect on power up-rating [75]. The release of fission gases also opens pathways for the release of other fission products and the release fraction of the fission gases can act as an easily measured indication for the release of volatile fission products to be the fuel-clad

the amount of phonon scattering; phonons providing the main contribution to thermal conductivity for fuel temperatures below 1600°C [66,70,71]. There may also be effects on the electronic thermal conduction processes, which dominate at higher temperatures, but the interactions are complex and will be intimately connected with effects arising from the fuel deviation from stoichiometry [72]. There is still much work to be done in this area if fuel burn-up limits are to be extended. Also affecting fuel temperature and fission product distribution is the radial power variation and a local increase in Pu concentration near the pellet surface in water reactor fuel. This process arises from neutron self-shielding by fissile atoms, accentuated by enhanced resonant capture of epithermal neutrons at the fuel surface. This leads to a reduction of both temperature and fission product release, for a given mean pin power [73]. However, there will be a locally enhanced concentration of fission products at the fuel surface from the enhanced fission rates. In conventional fuels the effect builds up during the fuel life; local powers reaching over a factor of two greater than the mean at 5% burn-up. In fuels with Gd,O, doping, as a soluble burnable poison, the local power variations are present from the start of life [73]. Interest in the behaviour of fission gases has increased over the last five years, as a result of the observation of anomalously high gas release values in fuel subjected to relatively mild power increases [74]. Fission gases are important because when released they can affect fuel life by pressurising the pin, this affects fuel temperatures and, as a consequence of the release o

offuel behaviour

potential

PWR

--__ --__

ramped ++

--_j --__

--__ 0 5465

--__ --__ High

oxygen

Annealed

potential

--__

-.__

--_

I 0.54601 0

I 10

I 20

I 30 Burn-up

40

I 50

60

(GWdltU)

Fig. 8. Variation of UO, cell parameter with bum-up for irradiated LWR and AGR fuel (measured by Pearce Windscale Nuclear Laboratory).

142

J. H. Gitttu et al. / Sajety aspects

60

Gas

release

at top

of ramp

offuel

behaviour

(%)

‘\

Matzke’s

ditfusmn

and

1

10 Temperature

Fig. 9. Comparison

of observed

burst

release

of fission

ramp

Small’s

100 rate

data

punts

IC

(C/s)

gas with observed releases in annealing experiments (measured by Small. Harwell Laboratory).

metal ratio [77]. This would be explained by a non-equilibrium state of the fission products in low temperature fuel that is maintained by fission damage processes. On heating, either in laboratory annealing experiments or a power ramp, the fuel is seen to experience a decrease in lattice parameter indicating an increasing oxygen to metal ratio. This could be a consequence of fission product precipitation and the attainment of thermodynamic equilibrium structures, resulting in a release of oxygen that was previously bound to fission products in solution, as was mentioned previously [46]. Local and possibly transient increases in oxygen content would increase the gas diffusion coefficients [80,81]. The in-pile observations of enhanced gas release in transients have now been supplemented with careful laboratory heating experiments, which enable the release process to be continuously monitored [82,83]. These experiments confirm the higher releases but additionally at heating rates of 12S”C/s burst releases have been observed during heating, which are substantially greater than expected using the enhanced diffusion rates, see fig. 9 [83]. These experiments and their accompanying microstructural examinations indicate that the processes underlying fission gas release are complicated and involve interaction between gas atoms and bubbles both within the fuel grains and on the grain boundaries [84]. The use of the mixed oxide fuel (MOX), consisting of a solid solution of plutonia and Urania, in a PWR has consequences on the fission gas behaviour. The diffusivity of fission gas atoms in MOX fuel is thought to be

similar to that of UOz, but the existence of small regions of high PuO, concentration in fuel manufactured by mechanically blending powders of UO, and PuO, results in higher overall release rates [85]. MOX fuel also produces more helium gas than UO,, as a result of the decay of higher actinides, in addition to helium formed by ternary fission [86]. The higher mobility of helium in the fuel can result in the early opening of interlinked pathways for release on the fuel grain boundaries. Perhaps the most important aspect of PWR transient operation is the problem of pellet-clad mechanical interaction (PCI). During a power change the fuel expands relative to the cladding as its temperature increases. When the fuel-clad gap is closed this results in a tensile load on the cladding. The problem is exacerbated by the generation of stress concentrations in the cladding. These arise from two sources. The fuel pellets themselves crack as a result of thermal stress generated by the radial temperature profile. Increasing the power leads to opening of the cracks as the pellet expands. This crack opening can produce a stress concentration in the clad, because of frictional resistance at the interface [87]. The other mechanism arises from the distortion of the pellets by the radial temperature profile which results in a wheatsheaf or hour-glass shape, which can be further aggravated by axial loading effects [88]. This pellet distortion can lead to ridging of the cladding but interfacial friction can also lead to stress concentrations in this case. The degree of stress concentration from these mechanisms will be dependent on

J. H. Girtus et al. / Safety aspects of fuel behaviour

the chemical state of the interface between the fuel and clad. Friction coefficients have been measured on unirradiated UO,-Zircaloy interfaces [89] but little is known of the effect of fission products and cladding corrosion products. Failure by PC1 is thought to be a result of stress corrosion cracking, induced by the fission product iodine accumulating in the fuel-clad gap. As is usual with stress corrosion cracking some threshold level of loading is required to nucleate the initial crack, which can then grow at lower stress [90,91]. The practical effect of this is that limits have to be set on the operating rating and the size of power increments applied to water reactor fuel. This has led to the development of smaller lower power fuel rod designs. A number of measures have been suggested to eliminate this problem. These include: modification to the pellet, such as chamfering or use of low density fuel; use of vibratory compacted spherical particle fuels; and the use of barrier layers of pure Zr or graphite at the fuel-clad interface.

Table 2 Melting sequence and pattern of the degradation of a PWR core Temperature of cladding surface (“C)

uo,+u (2 phase)

a-Zr( 0), (2 phase)

U-Zr

a-Zr(O), (2 phase)

Zircaloy

Phenomenon

700-750

Borosilicate starts to soften. Incipient melting of Ag-Cd-In alloy inside control rods

800

Fuel rods start to swell and burst and release some fission product species and fuel to the primary circuit

900

Reaction between Zr and steam starts. Temperatures of fuel rods rise more rapidly

1300-1500

First visible liquid formation due to (1) Inconel grid-Zircaloy eutectic reaction, and (2) UO,-Zircaloy reaction or exposed internal clad surfaces. Reaction 2 can result in vaporisation of fission product species

Above 1500

Further heating of exposed fuel results in the loss due to vaporisation of the volatile fission product species

1850-1950

Zircaloy starts to melt; subsequent behaviour is governed by thickness of ZrO, layer. Thin oxide layer - Zr melting and Zr + UO, reaction. Thick oxide layer-internal Zr surface may be protective and melting delayed

2.3. The behaviour of failed fueI in accidents If the Zircaloy cladding were to fail in fault conditions then those fission product elements present in the fuel-clad gap may enter the primary coolant. Much attention has been given to the fate of iodine, particularly of 13’1 in such faults. Of particular concern has been the possible revolatilisation of the iodide anion by radiolytic and thermal oxidation in loss of coolant faults [92]. When the capacity to cool the reactor core is lost then the fuel will overheat and, without restoration of the cooling, the fuel rods will degrade. The sequence of reactions in terms of temperatures for the various reactions leading to core degradation are shown in table 2. The reaction between unoxidised Zircaloy and uranium dioxide occurs at 1300-15OOOC. In the absence of direct contact between UO, and Zircaloy, oxygen can only be transported from the fuel to the cladding through the gas phase. Without externally applied pressure, no reaction has been observed between UO, and Zircaloy at temperatures up to 1500°C since there was no physical contact between pellets and Zircaloy cladding of fuel rods; the minimum solidus temperatures in the U-Zr-0 ternary phase diagram are at ca. 1300° C [93]. A detailed study of the reactions between UO, and Zircaloy has been reported by Hofmann and Kerwin-Peck [94]. The sequence of reaction layers at the interface between the reactants which is observed at room temperature is:

143

2400-2650

a)

ZrO, and UO,-ZrO,

mixtures melt

a) At the highest temperatures vaporisation of the less volatile fission product species has to be considered.

where cr-Zr(O), and cY-Zr(O), are ar-Zr with dissolved oxygen in two different two-phase regions separated by a U-Zr alloy single phase region. In general the number of reaction layers and their sequence is the same for all temperatures and reaction times. A very elegant analysis of the UO,-Zr couple in terms of the U-Zr-0 phase diagram and the transport properties of the system has been given by Olander [95]. The analysis was carried out for 1500 o C and accounted for the uranium rich alloy sandwiched between the two a-Zr(0) layers, the sequence of phases and the kinetics of layer growth. The diffusion path is shown in fig. 10. Some further work on the phase relationships is required particularly on the UO,-ZrO, section of the

144

J. H. Gittus et al. / Safety aspects

Fig. 10. U-Zr-0

offuelbehaomur

phase diagram at 1500 o C, after Olander [95].

phase diagram. Some preliminary attempts have been made to model these phase relationships by Imoto [96] together with some further experimental studies of the phase diagram by Yamanaka et al. [97,98]. A significant feature of these reactions is that an appreciable amount of UO,, up to 9 ~01% [94], can be dissolved by Zircaloy-4. In a fuel rod complete release of the volatile fission product from the molten fuel region can be expected. The presence of simulated fission product elements Cs, I and Te did not significantly influence the reaction of UO, and Zircaloy, although Te had reacted with the Sn of the Zircaloy [93]. The chemical form of the released fission product elements in various mixtures of hydrogen and steam is of great importance in determining their behaviour during the degradation of the core. The predominant caesium containing gaseous species are CsOH, CsI. The quantities of elemental I gas and HI gas in the reactor vessel and in some regions of the reactor coolant circuit

could be quite small. The chemical form of tellurium within the fuel pin would most probably be Cs,Te gas but this species would be unstable in mixtures of hydrogen and steam and than gaseous species such as Te and H,Te could appear. There has been considerable interest in the behaviour of Te in these accidents because in some conditions Te(g) could react with unoxidised Zircaloy cladding. However, when considerable oxidation of the cladding has occurred, the component of Sn present in this alloy concentrates in the metallic phase and can appear in the gaseous phase as SnTe. The caesium and tellurium containing phases can all react with or condense on surfaces and also form considerable quantities of material in the form of aerosol particles. The behaviour of caesium containing species has been a significant feature of many studies [99,100]. The quantification of the removal processes for the volatile fission products is an important part of the research programme on severe accident analysis. Tellurium gas

J. H. Gittus et al. / Safety aspects of fuel behaviour will also react with the stainless steel surfaces of the reactor coolant system and there have been studies on the kinetics of the deposition processes. In addition to the possible reactions of the fission product species with the walls of the reactor coolant system, a considerable amount of aerosol material will be formed due to the vaporization of Cd from the control rods, which are used in many pressurised water reactors, and consist of an alloy of cadmium, indium and silver clad in stainless steel. There is a considerable effort studying the behaviour of these aerosol materials in the presence of the volatile fission product species. The behaviour of the fission product species in the reactor coolant system is of importance, not only in the quantification of retention of material in the reactor coolant system, but also reactions within this circuit can modify the chemical forms of the fission product elements which are released from the degrading core before they enter the reactor containment. One such example is the conversion of CsI to HI by reaction with boric acid (101); the boric acid is present in the primary coolant. As the core heats up by the decay heat of the fission product nuclides and by the enthalpy of the reaction between Zircaloy and steam, ultimately a debris will be obtained which would consist of two immiscible liquids. These immiscible liquids derive from the behaviour of the U-Zr-0 system (fig. 10) and the chemical state of the debris including the structural and control rod material would be: (1) oxide phase containing Zr, U, Lanthanides, actinides, Ba, Sr, Nb, MO, and (2) metallic phase containing Zr, MO, Tc, Ru, Rh, Pd, Cr, Fe, Ni, Sn, Ag, In and some Te. As, Se, Sb. The exact distribution of the elements between the two phases would depend on the thermodynamic oxygen potential and the overall composition of the debris. Vaporization of the fission product species will result in loss of the actinides, lanthanides, Ba, Sr, and Ru from these mixtures of core debris. The most volatile species of these elements would be the hydroxides. Experimental determinations of the thermodynamic quantities of gaseous hydroxides are required; the only information on many of these systems is from assessment [102,103]. The predictions of the chemical constitution of core debris have been essentially supported by the detailed analysis of the debris from the TMI-2 reactor core. An apparent discrepancy has been the retention of caesium in the oxide phase of the debris. There are a number of gaseous molecules such as Cs,MoO, [65], Cs,CrO, [104,105] which could contribute to the releases of Cs and MO; evidence has also been

145

presented for the existence of the gaseous molecule CszTeO, [106]. It is important to ensure that all major species which could contribute to the loss of fission product elements from the core debris are characterised and their thermodynamic properties measured. 2.4. The interaction of core debris and concrete If the debris formed from the core materials were to melt through the reactor pressure vessel, reactions of this debris with the concrete of the containment would occur [107]. The chemical changes which occur during such reactions are the decomposition of the concrete giving both water vapour and carbon dioxide [log]. The sparging or streaming of these gases through the molten material can result in the formation of aerosols from the debris. These reactions could be an important source of radio-nuclides to the cavity and containment atmospheres during a severe accident in a light water reactor [109-1111. As the bubbles rise through the molten debris, the gas within them will undergo a change in composition due to reduction of CO,(gas) and H,O(gas) by some of the metals (especially any unoxidised Zr) and some gaseous species which evaporate from the surface of the bubbles will also be found within these bubbles. These latter species could be the main source for the formation of aerosol material; when the bubbles break through the surface of the pool many of the gaseous species will condense to form aerosol particles. We are endeavouring to estimate the amount of aerosol material which could be formed and its chemical constitution. The molten debris, like that which would be present before any melting through the stainless steel reactor pressure vessel, would consist of an oxide phase and a metal phase. The configuration of the melt can influence the composition of the vapour within the bubbles. Experimental studies and the modelling of systems which simulated such interactions are features of materials programmes for degraded core accidents.

3. Fast reactor fuel behaviour We will now investigate the safety aspects of the behaviour of fast reactor fuel. The fuel for fast reactors consists of a solid solution of 15-308 plutonia in Urania clad in stainless steel. The fuel is typically taken to a bum-up of 12% of the fissile atoms, but fuels are being developed to go to 20% fissile atom burn-up; this compares with typical burn-ups of up to 4% of the fissile atoms in water reactor fuel.

J. H. Gittus et al. / Sajety aspects

146

The high fuel bum-up coupled with high fuel centre temperatures, typically over 2000 o C from a relatively high coolant temperature and linear ratings around 45 kW/m, result in the almost complete release of fission gases from the fuel and substantial redistribution of the more volatile fission products. The hotter fuel and the non-oxidizing coolant produce fuel behaviour both in normal operation and in accidents, which differs in many respects from thermal reactor fuel. We will examine in this section: (9 The chemical constitution of mixed oxide fuel during operation at the higher temperatures associated with fast reactor operation. (ii) The specific aspects of fuel performance that distinguish fast reactor from water reactor fuel. (iii) The behaviour of failed fast reactor fuel and the constitution of the fuel after accidents in the presence of sodium. (iv) The properties of Urania and mixed oxide fuel at very high temperature, with reference to the effect of fission products wherever possible. 3.1. The chemical

constitution

offuel behauiour

shown in fig. 11 [34]. Considerable attention has been given to the measurement of the variation of thermodynamic oxygen potential with composition of the solid solution and temperature: the measurements are essentially confined to solutions with Pu : U ratios of technological significance. Most of these measurements [112-1241 have been fitted to a simple model by Lindemer and Besmann [35,36]; the phase is represented by a solid solution in which the components are Pu,,,O,, PuO,, UO, and either U,O,. 5 and U,O,. There are also some isolated defect models [16,31,38] as well as more detailed statistical thermodynamic models [13,25] which describe the thermodynamics of this phase. Measurements of thermodynamic oxygen potential have suggested that this quantity depends only on the average valency of the plutonium cation and not on the Pu concentration for the range of concentration appropriate to technological application of hypostoichiometric oxide and on the average valency of uranium for hyperstoichiometric oxide [112,113]. We shall see that thermodynamic oxygen potential is an important quantity in determining the behaviour of the oxide fuel. The quantities which are determined by oxygen potential are the diffusion coefficients of the actinide cations, which in turn determines the thermal creep behaviour of the fuel, the diffusion coefficients of the oxygen anions and those of the fission product atoms and ions in the lattice. The nature of the gaseous species above the solid phase is also determined by the oxygen potential. Redistribution of oxygen and the actinides occurs in the steep temperature gradients characteristic of oxide fuel. In hypostoichiometric fuel oxygen will diffuse down the temperature gradient to the colder regions of the fuel and the oxygen potential

of oxide fuel for fast reac-

tot-s

The fast reactor fuel is a solid solution of Urania and plutonia (with Pu: (U + Pu) ratios of 0.15-0.40) and oxygen anion : (Pu + U) cation ratios of 1.97-2.00. In the U-Pu-0 system there is an appreciable range of composition of the fluorite-structured solid solution of mania and plutonia in which U4+ cations can be oxidised or reduced and the Pu4+ cations reduced. The > reduction of U4+ cations occurs at temperatures 1500°C. The phase relationships in this system are

smblc M@,, Orthorhombic

M308_y + hl40g

‘\

I

--__

Mole

Part of a section through

the ternary

% Pu

phase diagram

for the U-Pu-0

systems between

400 and 800 o C.

147

J. H. Giitw et al. / Safety aspects of fuel behaviour

in the fuel-clad gap region will be of great significance in determining the compatibility between fuel and cladding. During irradiation of the fuel, as in thermal reactor fuel, the average valency of the fission products will be lower than that of the fuel and thus irradiation will result in an increase of oxygen potential in the fuel matrix. We have already indicated in our discussions of thermal reactor fuel that for fuel operating at ratings, the centre temperatures of which are sufficiently high for significant release of the volatile fission products to the fuel-clad gap or plenum of the rod to occur, that nucleation of additional phases within the matrix of the fuel will take place. Most of our considerable knowledge of the chemical effects of bum-up derive from post-irradiation examination of oxide fuels for all reactor systems, but the largest amount has come from the examination of fast reactor fuel. The formulation for the thermodynamic treatment of the multicomponent system was developed some twenty years ago [40] and has been much exploited and developed since then, Kleykamp [66] has provided an excellent review of the chemical state of the fission product elements in oxide fuels. Most of the examples presented considered in detail are taken from studies on fast reactor fuel. These examinations have allowed the following classifications of the fission product elements (from Se to Gd): Volatile fission product elements: Kr, Xe, Rb, Cs, Br, I, Se and Te. Fission product elements which form metallic phases: MO, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn and Sb. Fission product elements which form oxide phases: Rb, Cs, Sr, Ba, Zr, Nb, MO. Fission product elements which dissolve as oxides in the fuel matrix: Sr, Zr, Nb and Y, La, Ce, Pr, Nd, Pm and Sm. Molybdenum can be found both in an alloy phase and an oxide phase. Some of the fission product elements are found distributed between the fuel matrix and the separate oxide phase. The chemical state of the fission product elements in irradiated Urania is summarised in table 3. Post irradiation examination of fast reactor oxide fuels has shown that the predominant phases are the ‘so-called’ white metallic inclusions containing MoTc-Ru-Rl-Pd and the ‘so-called’ grey ceramic phase which contains mainly the oxides of barium, molybdenum, uranium, plutonium and zirconium [125]. Much attention has been given to the analysis of the metallic alloy [126-1281. The composition of the inclusion varies markedly and depends on the fission yields,

Table 3 The chemical state of fission products in irradiated urania breeder and Urania-plutonia solid solution fuel Fission product elements

Likely chemical state

Kr, Xe

Elemental state

Y, La-Eu

and actinides

Oxides which dissolve in host ma-

trix Ba, Sr

Oxides which can dissolve to a limited extent in the fuel and also form separate phases;

Ba,_,Sr,[Zr,~,_,_.Mo,U,Pu,]O, Br, I

Single phase halide solution; cs i -XRbXBr, --yIy

Rb, Cs

cs ,_,Rb,Br,_,I, and compounds analogous to Cs,UO, CWO,.,. Puy)04

Se, Te

eg. (Cs, -,Rb,),(U,

Single phase chalcogenide tion;

and -y-

solu-

(Csr-,Rb,),Sr,-,Te, Zr, Nb

Some dissolution in host matrix, see also Ba, Sr group

MO, Tc, Ru, Rh, Pd

Usually single phase alloy, sometimes two phase. Some MO can oxidise to MOO, and also form a compound analogous to Cs,MoO, - (Cs, _,Rb,)MoO,

Ag, Cd, In, Sn, Sb

Fission yields low; alloyed

the oxygen potential of the fuel material, the temperature gradient within or rating of the fuel and the extent of bum-up, as the oxygen potential can reach that required for the oxidation of the Mo-component of the alloy to MOO,. The metallic phases mainly have a hexagonal structure [129,130] which can exist over a wide range of composition [131-1331 (fig. 12). The main components of the alloy are MO and Ru, and these elements can migrate from hotter to colder regions of the fuel, as can Pd. Volatile gas phases species such as the oxides and Cs molybdates and ruthenates could be responsible for transport of MO and Ru; Pd would transport in elemental form. Two phase alloys have been observed with relatively high molybdenum contents, and at very high oxygen potentials and high bum-ups with very small MO contents. The oxidised MO was assumed to dissolve in the fuel matrix [134]; the

J. H. Giftus et al. / Safety aspects

148 (Rh*Pd)

offuel

hehauiour

pseudo-ternary

Rh o.ePdo.5

A(U,

phase diagram

Pu. Ln)O,-AZrO,-AMoO,

whereA=Ba,-,-,.Sr,Cs,; 17OO~C at.%

Fig. 12. An isothermal (1700 o C) section through the ternary phase diagram of the system MO-Ru-Rh,,-Pd.,. Metallic phases found in fast breeder reactor fuel have compositions that fall in the c (hexagonal) phase region.

ratio of MO concentration in the alloys to that in the fuel matrix was used as an indicator of the local oxygen potential of the fuel which is based on the equilibrium: Mo(dissolved

in alloy) + 0, (in fuel)

* MOO, (dissolved

x, ,~
This representation has been projected onto an isothermal section of the BaUO,-BaZrO,-BaMoO, system [125] (fig. 13). The amount of BaMoO, in the phase increases with oxygen potential and the final oxygen anion : metal cation ratio of the fuel. The compositions of the precipitates in the oxide fuel are found within the cubic single phase field of the pseudo-ternary system. These ternary phases will form eutectic liquids with the fuel matrix. UO, and BaUO, possess a eutectic at ca. 70 mol% UO, and 2569 + 30 K, for UO, and SrUO, the eutectic is at ca. 765 mol% UO, and 2639 f 30 K [137]. A small amount of liquid could be formed at temperatures considerably lower than that of the melting point of UO, (3120 K) or of the UO,-PuO, solid solution. Adamson et al. [137] have also found that the soluble fission product elements make only small reductions in the solidus temperatures of Urania-plutonia solutions; the effects on Urania would be similar. For “0.75 Pu 0,25O2 at 10% burn-up of the actinide atoms, the reduction in solidus temperatures was ca. 22 K. Pd which is found in the metallic phase can frequently be found associated with other fission product elements such as Ag, Cd, Sn, Sb and Te in fast reactor fuel [66]. It should also be noted that U and Pu could

in fuel).

There is, in fact, a limited solubility of MOO, in the fuel matrix [135]. Most of the oxidised molybdenum enters the multi-component oxide precipitate - the grey phase replacing the uranium, zirconium and other ions in the lattice. Thus the composition of the metallic phase can give indications of the local oxygen potential [136] and temperature which existed in the fuel during irradiation. The composition of the oxide phase, which precipitates, has been discussed by Kleykamp et al. [66,125] for fast reactor fuel. The precipitates consist of phases derived from the oxides of uranium and plutonium with the fission product elements Ba, Sr, Cs, Zr and MO together with the lanthanides (Ln). Studies using electron probe microanalysis indicate a composition [Ba,_,_,Sr,Cs,](U, Pu, Ln, Zr, Mo)O,. The lanthanide components can only be identified in fuel at very high bum-ups. The amount of MO in this phase clearly depends on the oxygen potential and the final oxygen anion : metal cation ratio of the fuel; the concentration of MO can reach a few weight percents at high values of Go>. The composition of this phase has been presented by Kleykamp et al. [125] on an isothermal section of a

BaMoOzj [ (Ba,Sr.Cs)

MoOa 1 ,

FBI?. defect

o FBA,(OIM)o= .

FBR/O/Wo.

196-1

A HTR-trw.(O/U)o.

[ (Ba.Sr.Cs)(U.Pu.RE)O3 BaUOs

1

97

1.98-2.00 2.00

[ (Ba.Sr.Cs)ZrOs 1 BaZrOg

Fig. 13. The composition of (Ba, Sr, Cs)((U, Pu, Ln)-Zr-Mo)O, precipitates found in irradiated HTR and fast breeder reactor fuels with different O/M ratios projected onto an isothermal section of Ba UO,-BaZrO, at 1700 o C [125].

J. H. Gittus et al. / Safety aspects of fuel behaoiour

be associated with Ru, Rh and Pd as U,_,Pu, (Ru,_,_,Rh,Pd,), at very low oxygen potentials such as could be found for failed LMFBR fuel. It has already been noted that for fuel operating at high temperatures redistribution of the actinide elements can occur; both gas phase and solid state mechanisms can be responsible with an increasing contribution from solid state diffusion with increasing temperature. Much attention has been given to the vaporization behaviour of Urania, plutonia, and uraniaplutonia [40,138,139]; the main gas phase species of these systems are the molecules UO, UO,, UO,, PuO and PuO,. We indicated earlier that the yields of radionuclides depend on the actual fissile atom and neutron energy. In considering the changes in chemical constitution we note that the increase in oxygen potential of the oxide with bum-up for plutonium fission [140] will be greater than that for uranium fission [141]. There have been a significant number of studies on the possible mechanisms of cladding failure due to chemical interactions between fission product species and the stainless steel. The corrosion of the cladding which is caused will reduce the effective can wall thickness and consequently could lead to pin failure at relatively short life-times. Yates and Linekar [142] have recently reviewed information on the fuel claddingchemical interaction within fuel pins from the Dounreay Prototype Fast Reactor. The cladding was 20% cold worked M316 stainless steel and the Urania-plutonia fuel contained 30% plutonia with an oxygen: uranium + plutonium ratio of 1.99. The fuel pellets were mainly annular with a smear density of 80% of theoretical density. The nominal linear ratings were 40 kW/m although some pins had higher ratings. Fuel clad chemical interaction (FCCI) has occurred to a varying degree in all these PFR pins which were examined. The interaction was of the same general form in all the normally rated pins; the corrosion was seen as a combination of reaction with the matrix and grain boundaries of the steel. In general, the maximum depth of clad attack increases with the ‘burn-up’ of the fuel. The depth of corrosion was expressed as a linear relationship with bum-up; the depth of the corrosion was 110 pm at 15% bum-up. The width of the fuel-clad gap which essentially determines the temperature gradient in the gap also influences the extent of corrosion. It has been shown that FCC1 in moderately rated fuel pins (I 45 kW/m) is a result of oxidation of the cladding. The attack of the matrix is often accompanied by some intergranular penetration and the rate of corrosion is slow. There is however, an additional mechanism of

149

corrosion which is seen in highly rated fuel pins in which melting of the fuel has occurred; and which gives severe cracking of the cladding. The column of the fuel appears to have retamed a solid outer crust and the melting of the fuel had resulted in an increase in the width of the gap. The deep cracks in the cladding were predominantly intergranular along the inner circumference of the cladding and there was no evidence of any chemical reaction in the cracks, but in the region close to the grain boundary the elements caesium, tellurium and molybdenum were all present in significant amounts. The ratio of tellurium to caesium was greater than that of the fission yields; high tellurium concentrations are believed to play a role in this type of corrosion. We realise that there is a significant amount of information from observations on irradiated fuel pins which provides the possibility of assessing the limits of performance of oxide fuel. We have just taken one example of these data from our experience of this topic in the United Kingdom. The prediction of the lifetimes of fuel elements under different operating conditions and in power transients has been a major objective. We shall briefly discuss here chemical aspects of the mechanisms of fuel-cladding interactions which have received considerable attention. The postulated mechanisms of both intergranular and uniform or matrix corrosion are oxidative and the factors which determine whether the FCC1 is revealed as deep intergranular corrosion or relatively small extents of matrix or uniform corrosion and the extents of such corrosion have been examined. The most important of these factors are the thermodynamic potential of oxygen and its availability, the thermodynamic activity of Te, usually expressed as the Cs: Te ratio, and the availability of Te and Cs for any reaction. The carbon activity of the stainless steel could also influence these reactions and the temperatures in the fuel-clad gap. A specific combination of these conditions is most probably required. The rating of the fuel and the width of the fuel-clad gap, and bum-up will determine the chemical constitution and hence the extent of reaction with the clad. Although enormous progress has been made in the modification of the factors which cause corrosion of the cladding it is not yet possible to predict when and at which point the cladding of fuel pins will fail due to these chemical processes, it is, however, possible to recommend regimes for operation which minimise the occurrence of the failure of the cladding. In addition to oxygen, the more volatile fission product elements or species will be found in the fuel-clad gap or plenum. There would be considerable release of

150

J.H. Girtus et al. / Sajety aspects o/fuel behaoiour

the elements rubidium and caesium, bromine and iodine, selenium and tellurium to the gap in addition to the rare gases, krypton and xenon and possibly molybdenum would also be found there. 3.2. Mixed oxide fuel performance Although the materials and configuration are similar, there are many differences in the behaviour between PWR and fast reactor fuel, which are largely the consequence of a different philosophy of the fuel operation. Whereas PWR fuel is now designed to retain fission gases, fast reactor fuel is operated at high linear rating and provision is made to accommodate essentially complete fission gas release. The use of high linear ratings means that the fuel centre temperature is high and a large fraction of the fuel is hot enough to deform easily by creep. Fast reactor fuels also undergo extensive thermal restructuring. There remains, however, a rim of fuel which has a high retention of fission gas, where thermal processes are slow and the only restructuring is that produced by fission fragment damage. The desire to reduce fuel cycle costs for fast reactors has led to progressive increases in the target bum-up. Originally mixed oxide fast reactor fuel was designed for only 7.5% fissile atom burn-up. These designs were found to be capable of going to 12% burn-up and with the adoption of low swelling cladding alloys the burn-up limits have been extended to 20% [143]. Both fuel cycle and reactor capital cost considerations have made it necessary to continue to operate the fuel at linear ratings of 45 kW/m or more. The desire to have a long operating cycle has also led to the adoption of larger fuel pin designs with low Pu concentrations. Such fuel designs have a larger internal Pu breeding effect which results in slower depletion of Pu from the fuel and the maintenance of high powers throughout the life of the pin. This combination of high powers at high bum-ups has extended the requirement for properties measurements in irradiated fuel. One of the design requirements is the determination of the linear rating that will produce melting in the fuel so that operating limits can be set. This requires an understanding of the effect of bum-up on thermal conductivity, which is an issue we have already discussed in the context of the PWR. Measurements have been made on mixed-oxide fuel containing simulated fission products [71] and some measurements have been made directly on fast reactor fuel at 4% fissile atom bum-up [144]. The build-up of fission gas bubble volume in the fuel and the reconfiguration of the fuel as a result of solid fission product swelling are found to be as important as the degrada-

tion of the thermal conductivity by soluble fission products. Mechanical interaction between the fuel pellets and the cladding during power changes is also of concern in the design of fast reactor fuel. The relatively high fuel temperatures mean that fuel creep is an important factor in determining interaction stresses. Fast reactor fuel is constructed with relatively large free volumes within the pin to absorb solid fission product swelling at high burn-up. This free volume is partly provided by the initial fuel-clad gap, partly from fuel porosity and, in current French and UK designs, in a central void. In fuel with solid pellets initially, restructuring processes frequently redistribute fuel porosity to form a central void. Solid fission product swelling progressively fills this volume and PC1 stresses increase as the available free volume diminishes. When the free volume is completely used up fuel creep processes are no longer able to relieve PCI loads and the fuel will become very sensitive to power changes. The magnitude of solid fission product swelling is thus important in fuel design. The amount of solid fission product swelling depends on the chemical form of the fission products, but the largest contribution arises from caesium. Caesium is found to migrate from hotter fuel regions to cooler regions and consequently swelling peaks can occur towards the ends of the pin [145]. Fission product induced corrosion will also affect the response of the fuel to PCI. Intragranular attack of the cladding by caesium alone is not found to embrittle the cladding apart from any reduction of wall thickness. Mixtures of Cs and Te, however, have been found to strongly embrittle stainless steels [146]. It has been postulated that the effect is due to liquid metal embrittlement, where the contact of the crack tip with the embrittling agent reduces the energy required to propagate a crack [147]. These observations of embrittlement of cladding out-of-pile have been used to try to explain the so-called ‘fuel adjacency effect’ when irradiated cladding samples are found to have considerably lower strength and ductibility after contact with fuel compared to clean samples irradiated to the same neutron dose [148]. This fuel adjacency effect is not always seen [149] and it may be that particular conditions of fuel chemical state are required to produce the effect. The laboratory observations require Te : Cs ratios greater than 1 : 1 for embrittlement, which is considerably higher than the ratio of Te: Cs produced by fission. If a high oxygen potential exists at the fuel-clad interface most of the Cs could be tied up in compounds leaving the low levels of metallic Cs required for embrittlement. In UO, increasing the oxygen to metal ratio of the

J. H. Gittus et al. / Safety aspects of fuel behaviour

fuel above the stoichiometric composition leads to increases in the kinetics of various transport processes [150]. The diffusion coefficients for self diffusion in the lattice, on surfaces and on grain boundaries are all increased, as are fission gas atom diffusion coefficients. Processes dependent on diffusion such as creep and sintering are similarly made more rapid. The ability of UO, to exist in a single phase hyperstoichiometric composition is known to occur through the accommodation of oxygen interstitials in the structure [151]. The increase in self diffusion is thought to occur by the interaction of the oxygen interstitials with the Schottky defect and anion Frenkel defect processes leading to an increase in the cation vacancy concentration [152,153]. In mixed-oxide fuel the situation is more complex. The presence of Pu enables the fuel to exist in a single phase state for substantial hypostoichiometric compositions at high temperatures. The oxygen potentials used in the sintering atmospheres mean that a slightly hypostoichiometric composition is usually encountered in fresh fuel. The diffusive processes are slowed for mixed-oxides just below the stoichiometric composition as an excess of oxygen vacancies inhibits the formation of cation vacancies. Other effects compensate for this and a minimum cation diffusion rate and creep rate is frequently seen for fuel with an oxygen to metal ratio around 1.98 [154,155]. It has been suggested that the recovery of the self diffusion coefficient at low oxygen contents was due to the stimulation of the cation interstitial concentrations from interacting with the cation Frenkel process [152,153], but is is now generally accepted that the energy for this process is too large for the process to be effective [156]. An explanation of tracer diffusion of Pu by the cyclic motion of defects in a cluster of two Pu3+ ions with an anion vacancy [157], but this process does not permit creep or other mass transfer effects. The inclusion of a wider range of defect interactions in the mass action calculations together with the effect of Pu3+ ions on the energy of formation and migration of defects should offer the possibility of explaining the observations more satisfactorily [158]. The presence of fission products is also likely to influence the diffusion coefficient of the material and also the creep behaviour. The main effects are likely to arise from the alteration of the material defect structure for a given oxygen potential. The high bum-ups that fast reactor fuels are subjected to make this a topic of great relevance to fuel performance assessment. Only very limited experimental measurements are available on the creep of mixed oxide fuels containing simulated fission products but these show significant effects [71].

151

Further investigations are currently being undertaken at Harwell. 3.3. Failed fuel behaviour Of the phases, given in table 3, those that are found in the fuel-clad gap would include some based on the following compounds: Cs,M,O,, Cs,MO, and Cs,M,O,, (where M = U,_,Pu,), Cs,Te, Cs,MoO,, Cs,CrO,, Cs,CrO,, Cs,CrO, and CsI. On failure of the cladding any ingress of sodium would result in the decomposition of the Cs uranoplutonates and of the caesium chromates to form sodium uranoplutonate (Na,MO,) and NaCrO,. Na and Cs are completely miscible in the liquid state. CsI and Cs,Te would also dissolve in the sodium. These reactions determine the extent of contamination by these soluble fission products, and the reaction of liquid sodium with the fuel matrix could lead to contamination of the primary circuit by fuel in particulate form. There have been many studies of the reactions between liquid sodium and Urania-plutonia solid solution and between liquid sodium and Urania and some of the lanthanide oxides [159]. For the Urania-plutonia solid solution the overall reaction can be expressed by: 3( y - z)Na( I, 0 dissolved) + (2 +r>U,-X,PuX~O2-z(s> -+ (Y - z)Na&J-,ZPuxI04(s) + (2+z)U,-,,Pu,,O,~,(s). Measurements have been made of the threshold oxygen potential for this reaction which is given by that for the three-phase assemblage (Na(l), U, _x,P~,,02_y, Na,U, _XzPu,zO,). Some measurements of this parameter using a solid state EMF cell technique are shown in fig. 14. The value of Go2 also determines the amount of dissolved oxygen in the liquid sodium which at equilibrium is ca. 1 wppm [160]. We thus have the value of 0 anion : (U + Pu) cation ratio or valency of Pu in the solid solution and the concentration of oxygen in liquid sodium above which the reaction occurs; thus for reaction the oxygen can be in either the fuel or the coolant. The sodium-fuel compound has a much lower density per actinide atom than the oxide and the reaction can result in considerable swelling in the region of the failure. Our determinations of the threshold Pu valency from measurements of the lattice parameters of the oxide phase in equilibrium with liquid sodium and sodium uranoplutonate together with measurements of the ex-

J. H. Gitttu et al. / Safety aspects of fuel behavrour

r . . . ‘Jo.~Puw%+Na

I

I

800

O O O

UOz+Na

x x x

U0.,Pu0.302+Na

I

900

I

1000

Temperature/K

Fig. 14. The experimentally determined variation of Go2 with temperature for sodium-Urania and sodium-Urania-plutonia.

tent of swelling of oxide pellets on reaction with liquid sodium indicate that the threshold Pu valency could decrease with increase in temperature of reaction and with increase in Pu concentration [161]. The analysis of these data also indicates that further measurements of the relationship between the composition of urania-plutonia solid solution and its thermodynamic oxygen potential are necessary. The dissolution of the fission product cations will influence the extent and rate of the reaction of sodium with irradiated fuel [162]. This has important consequences on the evolution of the failure process in fast reactor fuel; controlling the rate of dilation of the pin, the possibility of secondary failures and the release of fuel fragments. These considerations will influence the decision as to when to discharge fast reactor fuel after the first detection of fission products in the coolant circuit. 3.4. High temperature properties for severe accident analysis The licensing requirements of fast reactors require the investigation of a wide range of hypothetical accident conditions, many of which include the possibility of fuel melting. Fuel melting itself is not necessarily a dangerous condition but the movement of molten fuel needs to be fully understood as it can lead to large reactivity changes. Fuel motion towards the core centre can add reactivity but ejection of fuel from the core can rapidly and effectively terminate a reactivity excursion.

The properties of the fuel just below the melting point are important in the timing of melting and fuel failure and the properties above the melting point are important in the establishment of vapour pressures which are the main driving force for fuel and fission product redistribution. The most severe power transients are terminated by the vapour pressure driven disassembly of the reactor core. To understand this extreme board to hypothetical accidents much effort has been devoted to the prediction and measurement of the thermodynamic quantities for oxide fuels for temperatures to 5000 K and above. One of the most interesting aspects of oxide fuel high temperature properties is the possibility of a low order phase transition, due to disorder of the anion lattice at a temperature well below the melting temperature. This transition is referred to as the ‘Bredig’ transition after the originator of the proposal of this mechanism to explain anomalies in the heat capacity of heated fluorites, which have the same structure as UO, [163]. An exponential rise in the enthalpy of UO, is observed for temperatures above 1500 K, which cannot be explained by dilational effects, excitations of the crystal field or anharmonic lattice vibrations [164,165]. An anomaly in the specific heat is seen at ca. 2500” C which is generally accepted as the transition temperature. High temperature neutron diffraction studies at Harwell have now enabled the disorder of the anion lattice to be measured directly for UO, and ThO, [166]. In UO, significant anion lattice disorder is seen to start at around 1500 K and rise sharply for temperatures above 2500 K. The effect is found to be similar in ThO, when scaled by its melting temperature. Similar specific heat anomalies are seen in PuO, and mixed oxide [167]. In addition to the heat capacity the Bredig transition is found to affect a wide range of properties including the lattice constant, the elastic moduli and a range of phonon dispersion effects; these properties have also been measured using the high temperature neutron diffraction facility at Harwell up to temperatures of 3000 K [166]. Additionally the processes on the anion lattice may produce effects on the cation lattice, notably diffusion. There is evidence of a pronounced softening of UO, for temperatures above the Bredig transition which may be related to enhanced diffusion [168]. The melting temperature of UO, and mixed-oxide fuel will be affected by the presence of fission products. Soluble fission products, such as the lanthanides, will tend to reduce the melting point and increase the separation between the solidus and liquidus temperatures. A similar effect is seen on adding PuO, to UO,. The observations on the effect of irradiation on melting

J.H. Gittus et al. / Safety aspects

temperature in UO, have been very variable; ranging from essentially no significant effect to a drop of 200 ’ C on going to 50 GWd/t [168]. Measurements on mixed oxide fuel to higher burn-up have been more consistent and a drop of 100° C over 100 GWd/t is probably realistic [169]. Other fission product phases may melt at lower temperatures, in a manner that is not significant to fuel reactivity behaviour in accidents, but this can sometimes make assessment of melting limits difficult. The properties of molten fuel still need further investigation and in particular there are no measurements yet on the effect of fission products and plutonium on the basic properties of molten UO,. Measurements have been made on the viscosity, specific heat, density and thermal conductivity of liquid UO, [170,171]. Perhaps the most controversial of these properties is thermal conductivity. Firstly Kim et al. reported a value of 11 W/mK [172]. Using a similar method Otter and Damien found a value of 8.5 W/mK [173]. Tasman employing a rather different laser technique found a lower value of 2.2 W/mK [174]. A re-assessment of all the published results resulted in the proposition of 5.5 + 1.0 W/mK as the value [175]. Since then new measurements of Tasman have confirmed the previous measurements of a lower value [176], however analysis of surface temperatures of pulsed laser heating of UO, are consistent with a value of 5.5 W/mK [177]. This is an important problem that needs early resolution. The upper limit to energy release in a fast reactor overpower transient is controlled by the time of core dispersal due to the build-up of vapour pressure [178]. To calculate these limits an equation of state (EOS) of the oxide fuel is required which links the important quantities of energy, temperature, pressure and volume. Considerable effort to this end has been expended [179-1821 using techniques of extrapolation from experimental measurements usually below the liquidus temperature. Accurate data for the vapour pressure of fuel are required at temperatures of up to ca. 5000 K. Conventional techniques cannot be employed at such high temperatures because of the lack of a material for containment which is compatible with the liquid oxide. Some measurements on the liquid have been carried out using a transpiration technique [183]; in these experiments the containment was tungsten and measurements were made in the temperature range 3175-3390 K; at the highest temperature the composition of the liquid at the end of the experiment was UO,,,. In order to overcome the problems of materials compatibility a number of non-stationary pulse heating techniques require high speed diagnostics and high spatial and tem-

offuel behaviour

153

poral resolution. Measurements have been made of vapour pressure up to ca. 5000 K using laser beam heating; both Nd-YAG and CO, lasers have been employed [176-1891. Recently a heating technique using a CO, laser has been used to measure the boiling point of liquid UO, as a function of pressure of an inert gas, xenon [190]. Electron beam heating was employed to determine the relationship between enthalpy and pressure [191-1931. Fission heating has also been used [194-1961. Most of these data are for the vaporisation of uranium dioxide, but there are also some data for Urania-plutonia solutions. Data have been published for the variation of pressure with temperature in the temperature interval (3700-4090 K) [185], for U0,RP~,,20,,96 and pressure-enthalpy relationships for U,,,Pu,.,,O, [192,193]. In order to convert the data for the dependence of pressure on enthalpy to that of pressure on temperature a knowledge of the specific heat (C,) of the liquid is required. Experimental determinations of the enthalpy content of liquid UO, have been made between the melting point (3120 K) and 3520 K [197], which gave a constant value for C,, (135.8 J K-’ mol-‘). There are estimates of the variation of Cr with temperature which indicate that it could decrease with temperature [195,198]. The relationship used by Breitung and Reil [195,196] to convert the pressure-enthalpy data was Cp = 149.39 - 5.91 X 10p3T

J K-’

mol-‘.

The vapour pressure data are given for the temperature interval 4800-7500 K. Breitung and Reil give as best estimate relation for pressure and temperature over the range 3120-8500 K log P/MPa

= 23.7989 + 0.1505 - 29605.5/(

T/K)

- 4.7583 log (T/K).

It is suggested [195,196] that the data for vapour pressure obtained using the laser heating technique in which a Nd-YAG laser was employed showed higher temperature discrepancy than those of the other data. Ohse et al. [30] indicated that there was a discrepancy between the vapour pressure obtained by using models for the variation of oxygen potential with temperature and those from the measurements using a Nd-YAG laser. The discrepancy between the calculated pressure and measured rate of evaporation into a vacuum is believed to be due to an enhanced rate of evaporation caused by thermionic emission, a process involving surface and space charge effects. Magi11 et al. have recently indicated that there could be a significant contribution from ionic species in the

J.H. Gittus et al. / Safety aspects of fuel hehaviour

154

gas phase above uranium dioxide; the enhancement is significant in the temperature range 2000-6000 K [199]. This significant ion contribution in the gas phase is due to the equality between the recently published value of the electron affinity (A) of the UO, molecule and the ionisation energy (I) of UO, (5.2 and 5.5 eV respectively). The contribution of these ions to the gas phase moves the position of the congruently vaporising composition (CVC) to lower ratios of 0: U, although there is still considerable discrepancy between the calculated position of the CVC and that measured experimentally. The calculated position of CVC depends markedly on the model used for the variation of co * with composition of the solid phase. There are several extrapolations of the vapour pressure data into the liquid region using models for the calculation of oxygen potentials for Urania and Urania-plutonia solid solutions. Ohse et al. [30] give the for liquid UO,, variation of Go1 with temperature based on an extrapolation and assuming equipartition of the enthalpy of fusion between the uranium and oxygen components. The relationship given is c,Z(UG,.,

3liquid)/J

mall’

0,

= - 707000 + 1387-(K); no variation of specific heat is included. The partial pressures of all the gaseous species above the system can be calculated using the data for the Gibbs energies of formation of the gaseous species and condensed phase. The total equilibrium vapour pressure is given by log P,,,/MPa

= - 2.717 - 20131/T(K)

+ 1.925 log T.

The predicted vapour pressures are close to those given by the equation representing the experimental data. A solution to the problem of the apportionment of the enthalpy and entropy of fusion between the partial molal quantities for uranium and oxygen for Urania and between uranium-plutonium and oxygen for uraniaplutonia solution has been accomplished by Green and Leibowitz [200]. The extrapolation is based on the model for the solid of Blackbum (151. From these data and the Gibbs energies of the condensed phases and gaseous species, the variation of total pressure with temperature and composition has been derived. The measured parameters must ideally be fitted to a self consistent set, which forms the equation of state of the nuclear fuel. An excellent survey of the data for the critical point of UO, up to 1979 has been compiled by Ohse et al. [201]. The methods used to predict these data were the Principle of Corresponding States [202], the Law of Rectilinear Diameters [203,204], the Theory

of Significant Structures [205] and a Perturbed Hard Sphere Model [182]. A detailed comparison of the application of these approaches to the prediction of the critical constants and the formation of equations of state has been described by Browning et al. [198]. Fischer [179] has recently included the treatment of non-congruent evaporation into a model using Significant Structures Theory; the predicted critical temperature is ca. 10500 K, but will depend on the 0: U ratio; this temperature is significantly higher than previous assessments. This methodology could be applied to the prediction of the critical parameters of the mixed oxide. The vapour pressure of most fission products is considerably higher than that of the oxide fuel at liquid fuel temperature. A suggestion was made some time ago that the upper bond to energy release in severe fast reactor accidents could be significantly reduced if the vapour pressure of fission products could be taken into account [206]. There was some debate on the efficiency of the process because of the kinetics of vaporisation in a mixture and the problem of nucleating vapour bubbles in the liquid. The measurements on the vapour pressure of UO, and mixed oxide fuel in the Annular Core Research Reactor at Sandia [196] have now been extended to irradiated fuel [207]. The possibility of very rapid pressurisation due to the presence of fission products has been demonstrated.

4. Conclusions We have surveyed some of the important chemical and physical aspects of oxide fuel behaviour in normal and accident conditions appropriate to pressurised water and fast breeder reactors. The state of our knowledge is such that we can understand and predict the behaviour of fuel under a variety of conditions; we have indeed advanced a great deal in the past twenty-five years. Future studies aimed at further understanding of the properties of fuel and the influence of the vast number of fission product properties on fuel behaviour will allow us to have more confidence in the safe exploration of nuclear fission for power production. It would be inappropriate not to mention the outstanding contribution of the Institute for Transuranium Elements to our knowledge of fuel behaviour. The scientists of the Institute have made significant advances to our knowledge of: (i) Phase equilibria and thermodynamic properties of oxide and advanced fuel, including irradiated fuel. (ii) Mechanisms of gas release from irradiated fuel. (iii) Physical properties, including thermal conductivity.

J.H. Gittuc et al. / Safety aspects

Measurement of vapour pressures up to 6000 K using laser heating techniques and in addition to the experimental programmes there has also been the development of models to describe in-pile behaviour of oxides and advanced fuels, and also many contributions to the development of equations of state of oxide fuels for fast reactor safety analysis. We have not been able to consider the subject of advanced fuel, the uranium-plutonium carbides and nitrides for application in fast reactors. The Institute has been at the forefront of their development.

(iv)

offuel

We are very grateful to Mr. R.G.J. Ball of Chemistry Division, Hatwell Laboratory, for assistance in preparation of this paper and the associated presentation at the Transuranium seminar.

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[311 [321

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I331 1341 [351 [361 [371 [381 [391 [ml [411 [421 [431 PI

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