- Email: [email protected]
Fuel research and basic aspects of fuel in-pile performance
120
Journal
FUEL RESEARCH Hj. MATZKE
AND BASIC ASPECTS
of Nuclear
Materials 166 (1989) 120-131 North-Holland. Amsterdam
OF FUEL IN-PILE PERFORMANCE
and H. BLANK
Commrssion of the European Communities, Joint Research Centre, Institute for Transuranium Elements, Postfach 2340, D-7500 Karlsruhe, Fed. Rep. Germany
The approach adopted in the European Institute for Transuranium Elements to study basic aspects of the in-pile performance of a wide range of nuclear fuels (e.g. oxides, carbides, nitrides etc. for LWRs and LMFBRs) is described. It consists of an “all in-house activity” of fuel fabrication, careful and very detailed characterization and post-irradiation examination of fuels at different bum-ups and following different irradiation conditions, parallel laboratory studies under “clean conditions” and supporting theoretical and modelhng work. The approach is illustrated with typical examples of results obtained in the past 25 years. Finally, more recent results of basic experiments related to fuel in-pile performance are given.
1. Introduction Ever since the first large scale international action of the European Institute for Transuranium Elements, i.e. the production of the 1200 (metallic) fuel elements for the core of the French Masurca reactor, in its first year of operation, basic aspects of fuel in-pile performance were a major part of the Institute’s interests and activities. The philosophy was, and largely is, that of an “all in-house activity”, i.e. fuel fabrication, post-irradiation examination and supporting parallel laboratory work to provide the necessary basic physical and chemical data as well as extensive modelling activities with different computer codes are all done under one roof. The types of fuel investigated changed from the early metallic version over mixed oxides for fast breeders to dense, advanced LMFBR fuels and a large parametric study of the series MC-M(C, N)-MN (continued in a nitride programme up to now) to high burn-up and transient tested LWR oxide fuels. The approach of parallel work of (i) fuel fabrication, (ii) detailed post-irradiation examination of fuel rods, extended by special instrumented capsule irradiations of the same fuel, (iii) laboratory investigations of aspects relevant to in-pile performance (thus for instance including studies on simulated burn-up fuel and of fuel behaviour in out-of-pile thermal gradients of the type existing in-pile) together with (iv) detailed modelling and code work has enabled a thorough understanding of fuel in-pile performance to be achieved. Typical examples are given to illustrate the degree of understanding that has been elaborated. These exam-
0022-3115/89/$03.50 0 Elsevier Science Publishers (North-Holland Physics Publishing Division)
ples include fuel restructuring, fission product redistribution and release, radial oxygen redistribution in LMFBR oxides and radiation enhanced kinetics due to the effects of fission. The examples given are both from previous work performed within the 25 years of existence of the Transuranium Institute and from recent unpublished investigations, in order to give an overview appropriate for the celebration of the 25th anniversary of the Institute. The work of different laboratories of the Institute is described, including those for fabrication of fuel and of fuel elements (K. Richter et al.), characterization of as-fabricated fuel (C. Sari et al.), post-irradiation examination (M. Coquerelle, I.L.F. Ray, C. Walker et al.), supporting irradiations and instrumented capsules (M. Campana et al.), theoretical evaluation and modelling work (K. Lassmann, C. Ronchi et al.) and supporting parallel laboratory investigations (Hj. Matzke et al.). The list of references gives credit to the contributions of the numerous colleagues involved. Additional relevant information is contained in other articles in this volume, e.g. in those describing the oxide fuel transient programme [l], the modelling activities [2] and the work on nitride fuel [3].
2. Types of fuel fabricated and/or
analysed in TU
Within its first year of existence, the European Institute for Transuranium Elements produced 1200 fuel elements for the Masurca reactor. The fuel was metallic: 74% U, 25% Pu, 1% Fe. Further fuel elements of e.g.
B.V.
Hj. Matrke, H. Blank / Fuel research and fuel in-pile performance
Pu-Al alloys were fabricated for various other reactors or critical assemblies. The above “all in-house activity”, however, was only applied to ceramic fuels. These were oxides (U,,,Pu,,,) 0 2px for fast breeder reactors, irradiated in the Dounreay fast reactor, with the main variables being the O/M-ratio (or deviation x from stoichiometry) and the fuel density. Subsequently, advanced LMFBR fuels were fabricated, irradiated and analysed: carbides (U, Pu)C, nitrides (U, Pu)N, carbonitrides (U, Pu)C,_,N, and oxicarbides (U, Pu)(C, 0), the irradiations being performed in the Dounreay fast reactor, in Rapsodie, KNK 2 or Phenix. For the past 8 years, light water reactor fuels from commercial reactors (hence not fabricated in the Institute) were also increasingly investigated in order to determine and understand basic mechanisms of importance for LWR fuel performance (UO, and MOX, i.e. UO, with added PuO,) at high bum-up up to - 60 GWd/t U and in power transients with linear ratings in the range of - 40 to 46 kW/m. Most of the fuel was conventionally produced by cold-pressing powders and subsequent sintering, but a gel-supported production route, vibrocompaction, covibration of a mixture of e.g. 80% UC and 20% PUN to produce a carbonitride (U,~,PU,,,)C,,,N,,~ or other modified production procedures were also used. For the latter, the direct pressing procedure for carbides developed by Richter et al. [4] is a good example. It avoids handling of MC (or generally MX with X = C and/or N) powders with their high affinity for oxygen. In the following, examples will be given for basic aspects of most of these fuels.
3. Research approach and philosophy The “all in-house approach” of fuel fabrication, characterisation, post-irradiation examination, PIE, supporting capsule irradiations and laboratory work to study “single effect phenomena under clean conditions” with simultaneous extensive theoretical and modelling work provides optimal feedbacks in a number of ways, and allows effective use of both time for R&D and available budget. For instance, PIE may reveal aspects in the performance of a given fuel which appear undesirable. This information can be fed back into fuel and pin specifications and may even lead to a modified fuel production route since expertise in all relevant areas exists “under one roof”. In addition, iterative feedback between irradiation results and modelling activities can stepwise ameliorate both fuel performance codes and
121
models as well as the fuel itself. Furthermore, the availability of adequately equipped laboratories enabling the measurements of basic data needed for modelling activities in order to understand PIE or fabrication results, yields an effective improvement of fuel and understanding of fuel performance. Examples for such basic data are results on self-diffusion, fission product release, vapour pressures, phase diagrams and other thermodynamic data such as oxygen potentials for oxides, surface energies, mechanical and fracture properties, etc. The “clean conditions” or “single effect phenomena” refer to measurements where only one (or a few well controlled) parameters are varied at a time. Examples are measurements at constant O/U-ratio in UO,, release measurements with well characterized specimens containing one or a carefully selected specified number of fission products only, the latter for instance being introduced by controlled ion implantation. Studies with SIMFUEL, i.e. carefully prepared specimens containing inactive chemically added fission products are amongst other examples of “clean condition” experiments. For a full understanding of fuel performance, a knowledge of the thermal, mechanical and chemical changes in the operating fuel is necessary. Fig. 1 shows fuel cross sections of three main classes of non-metallic fuels investigated at TU, a fast breeder oxide (U, Pu)O,-,, an advanced fast breeder fuel, (U, Pu)C, and LWR fuel UO,, transient tested at 48 kW/m at a bum-up of 36 GWd/t U. Very different restructuring is obvious and different fuel temperatures (temperature gradients) and fission rates were operative. Typically, an LMFBR oxide showing columnar grain growth and formation of a central void has central temperatures of - 2200 QC, whereas the (much thicker) LMFBR carbide and the LWR oxide fuel rods are operated with (end of life) central temperatures about 1000” C lower. Obviously, the final structure is due to the combined effects of temperature and irradiation. The working philosophy of the TU approach consequently asks for an understanding of both the separate effects of temperature (thermal gradients, thermal stresses and vaporisation, diffusion processes) and of irradiation (changes in chemistry due to ingrowth of fission products, mechanical stresses due to swelling, radiation damage and radiation enhanced diffusion) as well as the possible synergistic effects of both operating at the same time. This can be achieved by the combination of irradiation tests of both fuel rods and fuel in instrumented capsules, and by skilfully planned laboratory experiments. To illustrate the procedure, four examples are given: (i) Fuel restructuring is measured at TU on many as-produced fuels in a number of laboratory furnaces producing
HJ. Matzke,
H. Blmk
/ Fuel research and fuel in-pile perfnrmuncr
I
LMFBR oxide
LWR - UO2 35.6 GWdltU, transient test 475 W/cm
Fig
1. Cross
sections
of fuel rods:
fast
breeder
oxide
and carbide, Coquerelle).
and transient
tested
light
water
reactor
oxide
(courtesy
M.
123
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance
temperature gradients in the fuel pellets very similar to those existing in-pile [5]. In parallel, by isothermal anneals, possible thermal densification of the fuel is measured. (ii) fracture stress, fracture toughness, flow stress, stress relaxation etc. are measured on as-fabricated fuel pellets using a variety of laboratory set-ups [6-Q Comparison with PIE data helps to separate thermal and irradiation effects. (iii) Extensive laboratory experiments on diffusion properties (self-diffusion, chemical or inter-diffusion, etc.) were complemented by parallel experiments with identical diffusion couples in a RF furnace in a nuclear reactor using instrumented capsules, thus again separating irradiation effects from thermal effects [9-121. (iv) Vapour pressures of ceramic fuels up to their melting point (and beyond) have been and are being measured to provide thermodynamic data and to understand and predict the in-pile behaviour of these fuels under normal operation and under accident conditions [13,14]. Other supporting irradiations in instrumented capsules were performed to provide high bum-up fuel (12% burn-up), to measure swelling at constant temperatures and constant temperature gradients, etc. These few examples must suffice. Of course, close feedback with modelling activities is necessary to achieve an effective performance of capsule irradiations, to optimize laboratory work and to use the information from laboratory work and capsule irradiations for the understanding of the fuel behaviour under actual operation conditions. Another aspect of the working philosophy is the detailed characterisation of fuel both before and after irradiation, including different microscopic techniques covering a wide range of resolutions in order to quantitatively measure features from the multi-micron to the sub-nanometer size range. Grain sizes and pores are measured by automated quantitative image analysis and optical micrography; larger fission gas bubbles are detected by replica electron microscopy; small precipitates, small bubbles, dislocation line densities, etc. are determined by transmission electron microscopy. The shielded scanning electron microscope and electron microprobe deliver further microscopic information, e.g. on diffusion profiles of fission gas within individual grains of irradiated fuel [15]. The different pieces of equipment are described in some detail elsewhere in these Proceedings [ 11. In addition to yielding a quantitative understanding of basic phenomena and mechanisms of fuel in-pile performance, the working approach provides an effective feedback to fuel fabrication in order to improve fuel properties and design.
4. Typical examples demonstrating the current understanding of in-pile mechanisms and fuel performance The space available does not allow a comprehensive description of all well understood basic aspects of fuel in-pile performance for the different types of fuels of section 2. This is contained in different review articles or, for the advanced fuels, in a monograph [16]. Rather, some typical examples of results of irradiation experiments, laboratory work and modelling will be given to illustrate the type of interplay and feedback that can be achieved. Such results have been obtained in a number of laboratories, but because of the occasion of the 25th anniversary of the European Institute for Transuranium Elements, the examples are taken from work performed in the Institute. Both previously published and recent, unpublished results will be given. We start with an early example of laboratory work providing the basis for understanding in-pile oxide re-
) 2500 \
2000
LO
1.5
1800
T,’
5.0 lO’/T, K-’
Fig. 2. Experimentally determined velocities of lenticular voids during the first restructuring of UO, in an out-of-pile thermal gradient [17].
HJ. Matrke,
124
H. Blank
/ Fuel research
structuring. Fig. 2 shows the velocities of lenticular voids in a laboratory temperature gradient simulating the shape of typical in-pile gradients [5]. Different symbols in fig. 2 are for different specimens and gradients. The presentation of the data corresponds to the kinetics of migration of underpressurized pores (p < 2-y/r, where y is the surface energy). The observed enthalpy (480 kJ/mol) is close to the value of the heat of vaporization. Such and similar experiments proved the purely thermal origin of the initial restructuring of oxide fuels at high ratings, demonstrated the important effect of fission gas pressure decreasing pore velocities, and showed the relative unimportance of the phenomenon for gas release in highly rated oxide fuels [17,18]. The second example is that of fission gas bubbles, the formation of which is the main reason for fuel swelling. Because of their low solubility, fission gases, if present at high concentration, precipitate into small intragranular bubbles. Such bubbles have been observed
Fig. 3. Transmission
electron
micrograph:
UO, implanted
and fuel m-pile performance
very early in UO,, and they have also been observed in irradiated carbides and nitrides. In theoretical treatments of bubble behavior, the bubbles are generally assumed to be filled with gas atoms only. This is true for fuels ion-implanted with rare gases, but not (necessarily) for reactor irradiated fuels. In recent years, ion implantation work as so-called “single-effect study” using one or two elements only, has shown that other volatiles can form “bubbles” in UO, as well. This phenomenon was demonstrated for Rb, Te, Cs and I [19-221, and an example is shown in fig. 3 for Rb implantation in UO, where bubbles are produced by annealing at 1500 o C and lead to Rb droplets within the previous (high temperature) bubble volume at room temperature, i.e. for the conditions of observation. Consequently, a realistic treatment of “fission gas bubbles” should allow for the presence of other gaseous or liquid volatiles within the bubble. Bubbles are very frequently located at dislocation lines, and are very often found
with 5 X 1015 Rb ions/en?,
45 keV, annealed
at 1500 o C (221
Hi. Matrke, H. Blank / Fuel research and fuel in-pile performance
125
Fig. 4. Scanning electron micrograph of LWR UO, fuel showing grain boundary bubbles containing five-metal particles (courtesy P. Knappik and M. Coquerelle).
attached to precipitates of rare metal fission products, the so-called five-metal precipitates consisting of Pd-Rl-Ru-Tc-Mo. This is particularly true for grain boundary bubbles as shown in fig. 4 [23]. Grain boundary bubbles form whenever gas atoms arrive at the grain boundaries, and their venting is usually the rate-determining step in final gas release to the gap or to the plenum in the fuel. These large grain-boundary bubbles with a relatively low gas pressure react on local stresses in the operating fuel. Final release is due to bubble interlinkage, often leading to the formation of tunnels at grain edges, and venting often occurs upon cooling, possibly in connection with fracture or microcrack formation. Fig. 5 shows recent results of venting of pores in UO, filled with high pressure Ar(1000 bar). Venting occurs in individual bursts and much more gas is released on cooling than on heating. The temperature range in which bursts occur in UO, is rather narrow (1500 to 1000 K). For UN, the temperature range is much wider (2500 to - 300 K). Any detailed modelling of gas release should also allow for these differences and for the statistical nature of pore venting. Other aspects of gas release treated in similar investigations are radiation-enhanced re-solution of gas from traps or bubbles, thermal solubility of rare gases measured following autoclave anneals, grain boundary
sweeping and bubble diffusion. For lack of space, these cannot be explicitly treated here. Instead we will conclude the treatment of rare gas kinetics and behaviour with three other examples: (i) when increasing the resolution of electron microscopy to the degree of lattice plane imaging, Xe bubbles can be observed to be formed even at room temperature [24] in UO, without any annealing treatment (see black arrow in fig. 6), (ii) a number of other experiments (gas release measurements from doped UO,, channeling and blocking experiments [25]) and calculations [26] give evidence that the equilibrium lattice site of a Xe atom in UO, is a trivacancy (Schottky trio) (see fig. 7 which also indicates a possible migration mechanism for Xe). Basic work has thus provided an insight into all aspects of gas release from explaining integral measurements on fuel rods down to the site of individual Xe atoms. (iii) Recently, a very early observation [27] of increased Kr and Xe diffusion rates in U02+ 1 has regained interest: studies of burn-up effects on fission gas release have shown increased release at high bum-up. Basic work, in this case on ion-implanted SIMFUEL (see above and ref. [20]), showed that the presence of fission products alone does not increase diffusion rates of rare gases or volatile fission products [28] (see fig. 8). Rather dramatic changes (unexpected low temperature release) in the diffusional
HJ. Matzke, H. Blunk / Fuel research
coolmg 13KS-’
L
I
500
I
lloo
I
2000 1500 temperature, K
Fig. 5. Rare gas release (Ar) from pores in UO, for two heating and one cooling rate, the highest annealing temperature being 2300 K. At the low heating rate of 3 K/s, venting of individual pores is resolved as bursts of gas.
behaviour at higher bum-up as postulated in recent gas release work [29] where the temperatures were calculated rather than measured would therefore not be expected to be based on the presence of dissolved fission products. Also, the hot cell anneals of high burn-up LWR UO, performed at the Institute (301 did not show low temperature release. The obvious conclusion was that the calculated temperatures were too low, probably due to a neglected decrease in thermal conductivity with increasing burn-up. There is, however, a remaining possibility to affect fission gas diffusion in UO, at high burn-up. Fission is generally assumed to increase the O/M-ratio of oxide fuel. Pu fission is more oxidizing than U fission because of the larger amounts of rare metal fission products in Pu fission. High burn-up UO, has an increased burn-up in its cold outer rim due to pronounced Pu formation caused by neutron resonance capture. If the clad does not buffer the O/U-ratio to near 2.00, changes in Xe diffusion rates and in the equilibrium site of Xe are expected: the early results of ref. [27] show significantly
andfuel
in-pile performunce
increased diffusion rates of Xe in UO,,,, as compared with UO,, and calculations [26] indicate that the equilibrium site of Xe in UOz+x can be charged trivacanties, tetravacancies or even single cation vacancies. If this is so, a synergistic effect occurs, augmenting or modifying the predicted change in O/U-ratio. Recent ion implantation and channeling results [31] show that Rb, Te and Cs occupy lattice sites following implantation at room temperature, and transmission electron microscopy [22] proves that (part of) these volatile fission products form precipitates or “bubbles” during anneals at or above - 1200 o C. These changes in lattice location unavoidably cause further changes in the O/U-ratio. The effect of fission products previously thought to be “chemically inert” and thus not changing fuel chemistry has only recently been treated in modelling work [32]. This effect can safely be assumed to be significant for high burn-up UO, fuel, though the absolute values of changes in O/U-ratio are still being discussed and investigated. Two examples of radiation effects are given to further illustrate basic aspects of fuel performance: Fig. 9 shows in-pile, radiation enhanced diffusion of U and Pu. i.e. the rate-controlling species for matter transport in ceramic fuels: a temperature-independent athermal fission enhanced diffusion is observed below - 1000 o C in all ceramic fuels studied, if the experiments are performed in-pile. The measured D* values for UO, cannot be explained with conventional collision or thermal spike effects. However, a fast mobility of uranium interstitials in the pressure gradient of the spike away from its axis could help to explain the experimental data. The ratio of D * for UO, : UC : UN corresponds closely to that of their thermal conductivity, and is confirmed by in-pile creep data (e.g. ref. [33)). These measurements help thus to explain the differences in plasticity at the fuel rim (hence at the clad) during reactor operation between oxides, carbides and nitrides. Fig. 10 shows recent results [31] on the behaviour of radiation induced point defects and points out further interesting differences between (the largely ionic) UO, and (the largely metallic) UN. Point defect concentrations in radiation damaged (or quenched) specimens are usually deduced from changes in properties such as electrical resistivity or lattice parameter. Such measurements do not allow a unique attribution to which type of defect (e.g. U, 0 or N vacancies, etc.) is formed or dnnealed. Rutherford backscattering allows this distinction. Fig. 10 shows U defects only, in the form of peaks in the aligned spectra. The areas and widths of these peaks are measures for the number and location of displaced U atoms. Both UO, and UN were damaged
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance
127
Fig. 6. Lattice plane images in ion implanted UO, [24].
by ion implantation. In the case of UO, and high damage levels, only - 1 U defect per incoming ion survived the bombardment at room temperature; there-
neutral trivacancy
charged trivacancy
EJ
@EC,
possible sites for Xe
m ,L+
-, 9
I+
El q
0
U- vacancy 0 -vacancy Xe-atom
possible saddle point for Xe migration
Fig. 7. Equilibrium
lattice site of Xe in UO, migration path.
and
possible
fore, more than 99% of the defects formed recombined immediately at room temperature. The remaining defects did not leave the range profile of the bombarding ions, i.e. they moved at the most a few nm during implantation and at ambient temperature. The important instantaneous damage recovery is also the reason for the fact that lattice images such as those of fig. 6 can be made with ion implanted UO,, and that UO, as a nuclear fuel does not become amorphous (as many other ceramics do). In contrast to UO,, a fraction of the U defects are seen to migrate away from the damaged surface layer in UN, as is obvious from the “dechanneling knee” located at lower energies (smaller channel numbers) and thus deeper in the crystal than the damage peak (see upper scale in fig. 10). Such dechanneling knees are typical for metals for which is is known that defects are mobile below room temperature. For UN, the dechanneling knee shows that U defects (probably interstitials) are mobile at room temperature and migrate out of the ion-implanted damaged zone into the virgin crystal. Further channeling experiments at liquid N, temperature have shown that these U defects actually move below room temperature. In addition, as with
Hj. Matrke, H. Blunk / Fuel research and fuel in-pile perjormance
128
z-i10 g
Kr
-a
5.3~1dZ ,ons/cm’
.j
- -
ThOz
o
SIMFUEL
0.5 -
UO,, important instantaneous damage recovery occurs in UN as well to a large extent. These results explain that UN does not become amorphous either during irradiation and they also help explain the radiation-enhanced diffusion coefficients D* of fig. 9. To conclude this short list of examples, the direct measurement of radial oxygen profiles in LMFBR mixed oxide fuels and of the increase of the oxygen potential AG(0,) in such fuels is mentioned. A solid state galvanic micro-cell is used for these investigations. Very recently, the AG(0,) of (U, Pu)Oz fuel from the Phenix reactor with a very high burn-up of 11 at% was successfully measured [34]. The results indicated a linear increase of AG(0,) of the fuel (initial composition (U, sPuO z )O, 9X) up to 11 at% burn-up without saturation.
0 9
t a_p
e-6’
8’
“0’
100 I 5 9.10’1lals/crn*
i 5. Summary and concluding remarks
temperature, “C Fig. 8. Release of Rb. I and Kr from ion-implanted (AECL, 3 at% burn-up [2X].
--
2000 I
CA _E 2
SIMFUEL
1500 1300 1100 I I a
The approach of the “al] in-house” activity adopted at the European Institute for Transuranium Elements to understand the in-pile performance of nuclear fuels was described with a number of examples of experiments giving insight in basic aspects. Detailed careful characterization of as-produced fuel and of fuel irradiated to different burn-up with supporting laboratory and modelling work provided not only reliable data for
900 800 700 h I I I .A
GM 200 ,dr ,
radiatlon -10
-
thermally
activated
enhanced o”:
normalized
oxides
. UO,, *IU,,,
f/&s
. DJ,h,,lO,
IU,,,Pu,,lC,
o UC,
*
nitrides
0 UN
4 HJ,,,Pu,,)N
diffusion
oPV+
to ~=5~10’z
Py,4&,
carbides
I50 T,“C
Itl,,k~,,~E
* WY1 Wu,,,,P~,,,I UI
-20
Fig. 9. Radiation
enhanced
i
6
8 lo’-
ld”’ 21
23 lO?T,
“a
25 K-’
diffusion of U and Pu in the ceramic fuels UO,, (U, Pu)O,, UC, (U, Pu)C, UN, (U, Pu)N, U(C, N) and (U, Pu)(C, N). The data are normalized to a fission rate of 5 X 10” fissions/cm3 s.
H. Blank / Fuel research and fuel in-pile performance
Hj. Matrke,
a3
0.2 ,
1.4
0.1
0
depth below surface,pm 0.3 0.2 0.1 0
0.4
I
1.6
1.6
129
1:4 E,MeV
1:s
118
---I .
0 5110’~ ions/cm’
l
:I$
2~10”
ions/cm’
,,d
mi’ _ _cleaved
280
--_--___’
320
1
360
s,
1
4Ml
I
I
.---------
440
electropolished
390
43o
470 t channel number
Fig. 10. Rutherford backscattering spectra of UO, (left) and UN (right) single crystals implanted with high dose Xe-ions (s = surface peak due to surface atoms which are not shielded along atomic strings, d = damage peak due to displaced U atoms situated within the range of the implanted ions, d.k. = dechanneling knee due to migration of U-atoms into the undamaged crystal). Both scattering in a random direction and in a direction where a low index direction of the crystal is aligned with the He beam are shown (channeling mode).
basic physical and chemical properties but allowed also to define mechanisms and kinetics important for in-pile performance. Knowledge of these properties and of these mechanisms is indeed necessary to develop “tailor-made” fuels and to understand (and predict) the in-pile behaviour of nuclear fuels. A convincing example for fuel fabrication is the KWU process of oxidative sintering of UO,, which reduces sintering temperatures from 1700 to 1100 o C, reduces also sintering times, enables the use of cheap furnaces and sintering boats and saves energy as well, The process is based on fundamental measurements of U diffusion in UO*+, showing a largely enhanced U-mobility (and thus enhanced sintering) with increasing deviation x from stoichiometry [35]. The above knowledge is also of paramount importance for modelling work. Rather frequently, fuel performance codes use imcomplete, oversimplified or even physically incorrect models. This is often tolerable if interpolation is asked for within the range of data to which the model is fitted, but extrapolation beyond this range is likely to yield invalid results. In this sense, to have no model is better than to use an incorrect one. Also, sometimes
surprising new mechanisms are postulated based on single isolated observations without considering the existing knowledge on basic aspects of fuel in-pile performance. The example of postulated gas release at very low temperatures in transient-tested UO, (but based on too low calculated temperatures) has already been given above. (Many laboratory and hot cell anneals of UO, irradiated at low temperatures had shown that such unexpected releases do not occur.) Other examples are assumptions of high thermal solubilities of rare gases just because locally in the fuel absence of fission gas bubbles is observed, or the postulation of U,O, existing in high bum-up UO, (Any U,O, existing in UO, would be transformed into UO,+. at reactor operating temperatures and U,O, would not be stable during fission either). Equally, the seemingly straightforward “conclusion” that gas release from oxide fuel must be fast because the fuel is hyperstoichiometric UO,,, necessitates different pieces of basic work and knowledge before the implied relation can be accepted. It is true that Kr and Xe diffuse significantly faster in UO,,, than in UO, (27) but this observation was made for single gas atom diffusion at low gas concentrations. It is
130
HJ. Mairke,
H. Blank / Fuel research rend fuel in-prle performunce
also well known that gas release at higher burn-up is a complex process of those single gas atoms becoming trapped, precipitating into bubbles, being freed from the bubbles by radiation-enhanced re-solution, reaching grain boundaries after multiple precipitation-re-solution processes, or by sweeping or bubble diffusion, and then having to wait for grain boundary bubble venting by coalescence, tunnel formation, crack formation etc. Unless it is shown that all these processes are accelerated in UO,,,, or that single gas atom diffusion between traps and bubbles is rate-controlling for release (it is known that this is not generally the case for release from UO,), a simple comparison of rare gas diffusion coefficients will not be appropriate to predict release from hyperstoichiometric fuel. For lack of space, only a limited number of examples could be given to illustrate the current advanced state of understanding basic aspects of fuel in-pile performance. Most examples were for oxides, since these are the most frequently used fuels, and most examples were from work performed at the European Institute for Transuranium Elements since these proceedings feature its 25 th anniversary. The R&D approach of the Institute has helped to achieve a thorough understanding of the in-pile mechanisms of different nuclear fuels on e.g. densification and swelling, restructuring, release of fission gases and other volatile elements, thermodiffusion in the existing temperature gradient and radiation enhanced diffusion, mechanical and fracture behaviour, chemical behaviour of fission products etc. The results are used as in-put data in mechanistic fuel performance codes.
6. Outlook for future investigations The R&D approach described can easily be applied to promising future research on nuclear fuels. Particular areas that should be investigated in detail are (i) problems of spent nuclear fuel direct storage of LWR fuel (and possibly other existing fuel), (ii) for LWR’s: MOX fuel and high burn-up UO, fuel, (iii) for LMFBRs: nitride fuels, and (iv) for nuclear waste aspects: questions related to transmutation of higher actinides (minor actinides) in suitable nuclear fuels.
Acknowledgements The authors would like to thank the numerous colleagues of the Institute who contributed with experimental and theoretical work to increase the knowledge
on basic aspects of fuel in-pile performance. The different contributing laboratories and the responsible scientists are listed in the introduction of this article. Thanks are also due to a large number of scientists from other laboratories (KfK, Germany; CEA, France: AERE. Harwell; AECL, Canada; ANL, USA and others, as well as those mentioned in the list of references).
References [I] Hj. Matzke. H. Blank, M. Coquerelle,
K. Lassmann, I.L.F. Ray, C. Ronchi and C.T. Walker, in: these Proceedings (TM Symposium), J. Nucl. Mater. 166 (1989) 162. [2] K. Lassmann, C. Ron&i and G. Small, in: these Proceedings (TTT Symposium). J. Nucl. Mater. 166 (1989) 109. [3] H. Blank. K. Richter, M. Coquerelle, Hi. Matzke, M. Campana, C. Sari and I.L.F. Ray, in: these Proceedings (TlT Symposium), J. Nucl. Mater. 166 (1989) 92. G. Kramer, C. Sari and P. [41 K. Richter, J. Gueugnon, Werner, Nucl. Technol. 70 (1985) 4. 151 C. Sari, Euratom Report EUR-5812 EN (1978). [61 P. Werner and H. Blank, Nucl. Technol. 52 (1981) 73. and P. [71 H. Blank, R. Mandler, Hj. Matzke, J.L. Routbort Werner. Res. Mechanica 5 (1983) 137. PI Hj. Matzke, V. Meyritz and J.L. Routbort, J. Am. Ceram. Sot. 66 (1983) 183. [91 Hj. Matzke. J. Nucl. Mater. 114 (1983) 121. and Hj. Matzke, Solid State Ionics 12 [lOI D. Glasser-Leme (1984) 217. 1111 Hj. Matzke, Radiat. Eff. 75 (1983) 317. in Non-Stoichiometric Comu21 Hj. Matzke, in: Transport pounds, Eds. G. Simkovich and V.S. Stubican (Plenum, New York, 1984) p. 331. v31 R.W. Ohse. J.F. Babelot. C. Cercignani, J.P. Hiernaut, M. Hoch, G.J. Hyland and J. Magill, J. Nucl. Mater. 130 (1985) 165. u41 J.F. Babelot, R.W. Ohse and M. Hoch, J. Nucl. Mater. 137 (1986) 144. u51 C.T. Walker and K. Lassmann, J. Nucl. Mater. 138 (1986) 155. Science of Advanced LMFBR Fuels, A [I61 Hj. Matzke, Monograph on Solid State Physics, Chemistry and Technology of Carbides, Nitrides and Carbonitrides of Uranium and Plutonium (North Holland, Amsterdam, 1986). u71 C. Ronchi and C. Sari, J. Nucl. Mater. 50 (1974) 91. [181 C. Ronchi and Hj. Matzke, in: Fuel and Fuel Elements for Fast Reactors, Vol. I (IAEA, Vienna, 1974) p. 57. v91 Hj. Matzke and C. Ronchi, Europ. Appl. Res. Rep. 5 (1984) 1105. PI R.A. Verrall, Hj. Matzke, T. Ogawa and B.J.F. Palmer, Canada Report AECL-9475 (1986) p. 558. [21] Hj. Matzke, I.L.F. Ray and R.A. Verrall, Report Int. Atom Energy Agency, IWGFPT/27 (IAEA, Vienna. 1987) p. 183.
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance [22] I.L.F. Ray, H. Thiele and Hj. Matzke, unpublished results. [23] M. Mogensen and C.T. Walker, personal communication, presented at Enlarged Halden Meeting, Loen, May 1988. [24] Hj. Mat&e, C. Rossouw and S.E. Donnelly, unpublished results. [25] Hj. Matzke, Radiat. Eff. 53 (1980) 219. [26] R.A. Jackson and C.R.A. Catlow, J. Nucl. Mater. 127 (1985) 161, 167. [27] R. Lindner and Hj. Matzke, Z. Naturforschung 14a (1959) 582; AERE-Trans 845 and CEA-tr-A-856 (1959). 1281 Hj. Matzke, R.A. VerraII and B.J.F. Palmer, unpublished results.
[29] [30] [31] [32]
131
I. Misfeldt, J. Nucl. Mater. 135 (1985) 1. M. Coquerelle, personal communication. Hj. Matzke, S. Fritz and A. Turos, unpublished results. D.A. Mac Innes and P.W. Winter, J. Phys. Chem. Solids 49 (1988) 143. [33] D. Brucklacher and W. Dienst, J. Nucl. Mater. 42 (1972) 285. [34] Hj. Matzke, J. Ottaviani, D. Pellottiero an J. Rouault, J. Nucl. Mater. 160 (1988) 142. [35] Hj. Matzke, Euratom Report EUR-7700 DE (1982) p. 139.
Journal
FUEL RESEARCH Hj. MATZKE
AND BASIC ASPECTS
of Nuclear
Materials 166 (1989) 120-131 North-Holland. Amsterdam
OF FUEL IN-PILE PERFORMANCE
and H. BLANK
Commrssion of the European Communities, Joint Research Centre, Institute for Transuranium Elements, Postfach 2340, D-7500 Karlsruhe, Fed. Rep. Germany
The approach adopted in the European Institute for Transuranium Elements to study basic aspects of the in-pile performance of a wide range of nuclear fuels (e.g. oxides, carbides, nitrides etc. for LWRs and LMFBRs) is described. It consists of an “all in-house activity” of fuel fabrication, careful and very detailed characterization and post-irradiation examination of fuels at different bum-ups and following different irradiation conditions, parallel laboratory studies under “clean conditions” and supporting theoretical and modelhng work. The approach is illustrated with typical examples of results obtained in the past 25 years. Finally, more recent results of basic experiments related to fuel in-pile performance are given.
1. Introduction Ever since the first large scale international action of the European Institute for Transuranium Elements, i.e. the production of the 1200 (metallic) fuel elements for the core of the French Masurca reactor, in its first year of operation, basic aspects of fuel in-pile performance were a major part of the Institute’s interests and activities. The philosophy was, and largely is, that of an “all in-house activity”, i.e. fuel fabrication, post-irradiation examination and supporting parallel laboratory work to provide the necessary basic physical and chemical data as well as extensive modelling activities with different computer codes are all done under one roof. The types of fuel investigated changed from the early metallic version over mixed oxides for fast breeders to dense, advanced LMFBR fuels and a large parametric study of the series MC-M(C, N)-MN (continued in a nitride programme up to now) to high burn-up and transient tested LWR oxide fuels. The approach of parallel work of (i) fuel fabrication, (ii) detailed post-irradiation examination of fuel rods, extended by special instrumented capsule irradiations of the same fuel, (iii) laboratory investigations of aspects relevant to in-pile performance (thus for instance including studies on simulated burn-up fuel and of fuel behaviour in out-of-pile thermal gradients of the type existing in-pile) together with (iv) detailed modelling and code work has enabled a thorough understanding of fuel in-pile performance to be achieved. Typical examples are given to illustrate the degree of understanding that has been elaborated. These exam-
0022-3115/89/$03.50 0 Elsevier Science Publishers (North-Holland Physics Publishing Division)
ples include fuel restructuring, fission product redistribution and release, radial oxygen redistribution in LMFBR oxides and radiation enhanced kinetics due to the effects of fission. The examples given are both from previous work performed within the 25 years of existence of the Transuranium Institute and from recent unpublished investigations, in order to give an overview appropriate for the celebration of the 25th anniversary of the Institute. The work of different laboratories of the Institute is described, including those for fabrication of fuel and of fuel elements (K. Richter et al.), characterization of as-fabricated fuel (C. Sari et al.), post-irradiation examination (M. Coquerelle, I.L.F. Ray, C. Walker et al.), supporting irradiations and instrumented capsules (M. Campana et al.), theoretical evaluation and modelling work (K. Lassmann, C. Ronchi et al.) and supporting parallel laboratory investigations (Hj. Matzke et al.). The list of references gives credit to the contributions of the numerous colleagues involved. Additional relevant information is contained in other articles in this volume, e.g. in those describing the oxide fuel transient programme [l], the modelling activities [2] and the work on nitride fuel [3].
2. Types of fuel fabricated and/or
analysed in TU
Within its first year of existence, the European Institute for Transuranium Elements produced 1200 fuel elements for the Masurca reactor. The fuel was metallic: 74% U, 25% Pu, 1% Fe. Further fuel elements of e.g.
B.V.
Hj. Matrke, H. Blank / Fuel research and fuel in-pile performance
Pu-Al alloys were fabricated for various other reactors or critical assemblies. The above “all in-house activity”, however, was only applied to ceramic fuels. These were oxides (U,,,Pu,,,) 0 2px for fast breeder reactors, irradiated in the Dounreay fast reactor, with the main variables being the O/M-ratio (or deviation x from stoichiometry) and the fuel density. Subsequently, advanced LMFBR fuels were fabricated, irradiated and analysed: carbides (U, Pu)C, nitrides (U, Pu)N, carbonitrides (U, Pu)C,_,N, and oxicarbides (U, Pu)(C, 0), the irradiations being performed in the Dounreay fast reactor, in Rapsodie, KNK 2 or Phenix. For the past 8 years, light water reactor fuels from commercial reactors (hence not fabricated in the Institute) were also increasingly investigated in order to determine and understand basic mechanisms of importance for LWR fuel performance (UO, and MOX, i.e. UO, with added PuO,) at high bum-up up to - 60 GWd/t U and in power transients with linear ratings in the range of - 40 to 46 kW/m. Most of the fuel was conventionally produced by cold-pressing powders and subsequent sintering, but a gel-supported production route, vibrocompaction, covibration of a mixture of e.g. 80% UC and 20% PUN to produce a carbonitride (U,~,PU,,,)C,,,N,,~ or other modified production procedures were also used. For the latter, the direct pressing procedure for carbides developed by Richter et al. [4] is a good example. It avoids handling of MC (or generally MX with X = C and/or N) powders with their high affinity for oxygen. In the following, examples will be given for basic aspects of most of these fuels.
3. Research approach and philosophy The “all in-house approach” of fuel fabrication, characterisation, post-irradiation examination, PIE, supporting capsule irradiations and laboratory work to study “single effect phenomena under clean conditions” with simultaneous extensive theoretical and modelling work provides optimal feedbacks in a number of ways, and allows effective use of both time for R&D and available budget. For instance, PIE may reveal aspects in the performance of a given fuel which appear undesirable. This information can be fed back into fuel and pin specifications and may even lead to a modified fuel production route since expertise in all relevant areas exists “under one roof”. In addition, iterative feedback between irradiation results and modelling activities can stepwise ameliorate both fuel performance codes and
121
models as well as the fuel itself. Furthermore, the availability of adequately equipped laboratories enabling the measurements of basic data needed for modelling activities in order to understand PIE or fabrication results, yields an effective improvement of fuel and understanding of fuel performance. Examples for such basic data are results on self-diffusion, fission product release, vapour pressures, phase diagrams and other thermodynamic data such as oxygen potentials for oxides, surface energies, mechanical and fracture properties, etc. The “clean conditions” or “single effect phenomena” refer to measurements where only one (or a few well controlled) parameters are varied at a time. Examples are measurements at constant O/U-ratio in UO,, release measurements with well characterized specimens containing one or a carefully selected specified number of fission products only, the latter for instance being introduced by controlled ion implantation. Studies with SIMFUEL, i.e. carefully prepared specimens containing inactive chemically added fission products are amongst other examples of “clean condition” experiments. For a full understanding of fuel performance, a knowledge of the thermal, mechanical and chemical changes in the operating fuel is necessary. Fig. 1 shows fuel cross sections of three main classes of non-metallic fuels investigated at TU, a fast breeder oxide (U, Pu)O,-,, an advanced fast breeder fuel, (U, Pu)C, and LWR fuel UO,, transient tested at 48 kW/m at a bum-up of 36 GWd/t U. Very different restructuring is obvious and different fuel temperatures (temperature gradients) and fission rates were operative. Typically, an LMFBR oxide showing columnar grain growth and formation of a central void has central temperatures of - 2200 QC, whereas the (much thicker) LMFBR carbide and the LWR oxide fuel rods are operated with (end of life) central temperatures about 1000” C lower. Obviously, the final structure is due to the combined effects of temperature and irradiation. The working philosophy of the TU approach consequently asks for an understanding of both the separate effects of temperature (thermal gradients, thermal stresses and vaporisation, diffusion processes) and of irradiation (changes in chemistry due to ingrowth of fission products, mechanical stresses due to swelling, radiation damage and radiation enhanced diffusion) as well as the possible synergistic effects of both operating at the same time. This can be achieved by the combination of irradiation tests of both fuel rods and fuel in instrumented capsules, and by skilfully planned laboratory experiments. To illustrate the procedure, four examples are given: (i) Fuel restructuring is measured at TU on many as-produced fuels in a number of laboratory furnaces producing
HJ. Matzke,
H. Blmk
/ Fuel research and fuel in-pile perfnrmuncr
I
LMFBR oxide
LWR - UO2 35.6 GWdltU, transient test 475 W/cm
Fig
1. Cross
sections
of fuel rods:
fast
breeder
oxide
and carbide, Coquerelle).
and transient
tested
light
water
reactor
oxide
(courtesy
M.
123
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance
temperature gradients in the fuel pellets very similar to those existing in-pile [5]. In parallel, by isothermal anneals, possible thermal densification of the fuel is measured. (ii) fracture stress, fracture toughness, flow stress, stress relaxation etc. are measured on as-fabricated fuel pellets using a variety of laboratory set-ups [6-Q Comparison with PIE data helps to separate thermal and irradiation effects. (iii) Extensive laboratory experiments on diffusion properties (self-diffusion, chemical or inter-diffusion, etc.) were complemented by parallel experiments with identical diffusion couples in a RF furnace in a nuclear reactor using instrumented capsules, thus again separating irradiation effects from thermal effects [9-121. (iv) Vapour pressures of ceramic fuels up to their melting point (and beyond) have been and are being measured to provide thermodynamic data and to understand and predict the in-pile behaviour of these fuels under normal operation and under accident conditions [13,14]. Other supporting irradiations in instrumented capsules were performed to provide high bum-up fuel (12% burn-up), to measure swelling at constant temperatures and constant temperature gradients, etc. These few examples must suffice. Of course, close feedback with modelling activities is necessary to achieve an effective performance of capsule irradiations, to optimize laboratory work and to use the information from laboratory work and capsule irradiations for the understanding of the fuel behaviour under actual operation conditions. Another aspect of the working philosophy is the detailed characterisation of fuel both before and after irradiation, including different microscopic techniques covering a wide range of resolutions in order to quantitatively measure features from the multi-micron to the sub-nanometer size range. Grain sizes and pores are measured by automated quantitative image analysis and optical micrography; larger fission gas bubbles are detected by replica electron microscopy; small precipitates, small bubbles, dislocation line densities, etc. are determined by transmission electron microscopy. The shielded scanning electron microscope and electron microprobe deliver further microscopic information, e.g. on diffusion profiles of fission gas within individual grains of irradiated fuel [15]. The different pieces of equipment are described in some detail elsewhere in these Proceedings [ 11. In addition to yielding a quantitative understanding of basic phenomena and mechanisms of fuel in-pile performance, the working approach provides an effective feedback to fuel fabrication in order to improve fuel properties and design.
4. Typical examples demonstrating the current understanding of in-pile mechanisms and fuel performance The space available does not allow a comprehensive description of all well understood basic aspects of fuel in-pile performance for the different types of fuels of section 2. This is contained in different review articles or, for the advanced fuels, in a monograph [16]. Rather, some typical examples of results of irradiation experiments, laboratory work and modelling will be given to illustrate the type of interplay and feedback that can be achieved. Such results have been obtained in a number of laboratories, but because of the occasion of the 25th anniversary of the European Institute for Transuranium Elements, the examples are taken from work performed in the Institute. Both previously published and recent, unpublished results will be given. We start with an early example of laboratory work providing the basis for understanding in-pile oxide re-
) 2500 \
2000
LO
1.5
1800
T,’
5.0 lO’/T, K-’
Fig. 2. Experimentally determined velocities of lenticular voids during the first restructuring of UO, in an out-of-pile thermal gradient [17].
HJ. Matrke,
124
H. Blank
/ Fuel research
structuring. Fig. 2 shows the velocities of lenticular voids in a laboratory temperature gradient simulating the shape of typical in-pile gradients [5]. Different symbols in fig. 2 are for different specimens and gradients. The presentation of the data corresponds to the kinetics of migration of underpressurized pores (p < 2-y/r, where y is the surface energy). The observed enthalpy (480 kJ/mol) is close to the value of the heat of vaporization. Such and similar experiments proved the purely thermal origin of the initial restructuring of oxide fuels at high ratings, demonstrated the important effect of fission gas pressure decreasing pore velocities, and showed the relative unimportance of the phenomenon for gas release in highly rated oxide fuels [17,18]. The second example is that of fission gas bubbles, the formation of which is the main reason for fuel swelling. Because of their low solubility, fission gases, if present at high concentration, precipitate into small intragranular bubbles. Such bubbles have been observed
Fig. 3. Transmission
electron
micrograph:
UO, implanted
and fuel m-pile performance
very early in UO,, and they have also been observed in irradiated carbides and nitrides. In theoretical treatments of bubble behavior, the bubbles are generally assumed to be filled with gas atoms only. This is true for fuels ion-implanted with rare gases, but not (necessarily) for reactor irradiated fuels. In recent years, ion implantation work as so-called “single-effect study” using one or two elements only, has shown that other volatiles can form “bubbles” in UO, as well. This phenomenon was demonstrated for Rb, Te, Cs and I [19-221, and an example is shown in fig. 3 for Rb implantation in UO, where bubbles are produced by annealing at 1500 o C and lead to Rb droplets within the previous (high temperature) bubble volume at room temperature, i.e. for the conditions of observation. Consequently, a realistic treatment of “fission gas bubbles” should allow for the presence of other gaseous or liquid volatiles within the bubble. Bubbles are very frequently located at dislocation lines, and are very often found
with 5 X 1015 Rb ions/en?,
45 keV, annealed
at 1500 o C (221
Hi. Matrke, H. Blank / Fuel research and fuel in-pile performance
125
Fig. 4. Scanning electron micrograph of LWR UO, fuel showing grain boundary bubbles containing five-metal particles (courtesy P. Knappik and M. Coquerelle).
attached to precipitates of rare metal fission products, the so-called five-metal precipitates consisting of Pd-Rl-Ru-Tc-Mo. This is particularly true for grain boundary bubbles as shown in fig. 4 [23]. Grain boundary bubbles form whenever gas atoms arrive at the grain boundaries, and their venting is usually the rate-determining step in final gas release to the gap or to the plenum in the fuel. These large grain-boundary bubbles with a relatively low gas pressure react on local stresses in the operating fuel. Final release is due to bubble interlinkage, often leading to the formation of tunnels at grain edges, and venting often occurs upon cooling, possibly in connection with fracture or microcrack formation. Fig. 5 shows recent results of venting of pores in UO, filled with high pressure Ar(1000 bar). Venting occurs in individual bursts and much more gas is released on cooling than on heating. The temperature range in which bursts occur in UO, is rather narrow (1500 to 1000 K). For UN, the temperature range is much wider (2500 to - 300 K). Any detailed modelling of gas release should also allow for these differences and for the statistical nature of pore venting. Other aspects of gas release treated in similar investigations are radiation-enhanced re-solution of gas from traps or bubbles, thermal solubility of rare gases measured following autoclave anneals, grain boundary
sweeping and bubble diffusion. For lack of space, these cannot be explicitly treated here. Instead we will conclude the treatment of rare gas kinetics and behaviour with three other examples: (i) when increasing the resolution of electron microscopy to the degree of lattice plane imaging, Xe bubbles can be observed to be formed even at room temperature [24] in UO, without any annealing treatment (see black arrow in fig. 6), (ii) a number of other experiments (gas release measurements from doped UO,, channeling and blocking experiments [25]) and calculations [26] give evidence that the equilibrium lattice site of a Xe atom in UO, is a trivacancy (Schottky trio) (see fig. 7 which also indicates a possible migration mechanism for Xe). Basic work has thus provided an insight into all aspects of gas release from explaining integral measurements on fuel rods down to the site of individual Xe atoms. (iii) Recently, a very early observation [27] of increased Kr and Xe diffusion rates in U02+ 1 has regained interest: studies of burn-up effects on fission gas release have shown increased release at high bum-up. Basic work, in this case on ion-implanted SIMFUEL (see above and ref. [20]), showed that the presence of fission products alone does not increase diffusion rates of rare gases or volatile fission products [28] (see fig. 8). Rather dramatic changes (unexpected low temperature release) in the diffusional
HJ. Matzke, H. Blunk / Fuel research
coolmg 13KS-’
L
I
500
I
lloo
I
2000 1500 temperature, K
Fig. 5. Rare gas release (Ar) from pores in UO, for two heating and one cooling rate, the highest annealing temperature being 2300 K. At the low heating rate of 3 K/s, venting of individual pores is resolved as bursts of gas.
behaviour at higher bum-up as postulated in recent gas release work [29] where the temperatures were calculated rather than measured would therefore not be expected to be based on the presence of dissolved fission products. Also, the hot cell anneals of high burn-up LWR UO, performed at the Institute (301 did not show low temperature release. The obvious conclusion was that the calculated temperatures were too low, probably due to a neglected decrease in thermal conductivity with increasing burn-up. There is, however, a remaining possibility to affect fission gas diffusion in UO, at high burn-up. Fission is generally assumed to increase the O/M-ratio of oxide fuel. Pu fission is more oxidizing than U fission because of the larger amounts of rare metal fission products in Pu fission. High burn-up UO, has an increased burn-up in its cold outer rim due to pronounced Pu formation caused by neutron resonance capture. If the clad does not buffer the O/U-ratio to near 2.00, changes in Xe diffusion rates and in the equilibrium site of Xe are expected: the early results of ref. [27] show significantly
andfuel
in-pile performunce
increased diffusion rates of Xe in UO,,,, as compared with UO,, and calculations [26] indicate that the equilibrium site of Xe in UOz+x can be charged trivacanties, tetravacancies or even single cation vacancies. If this is so, a synergistic effect occurs, augmenting or modifying the predicted change in O/U-ratio. Recent ion implantation and channeling results [31] show that Rb, Te and Cs occupy lattice sites following implantation at room temperature, and transmission electron microscopy [22] proves that (part of) these volatile fission products form precipitates or “bubbles” during anneals at or above - 1200 o C. These changes in lattice location unavoidably cause further changes in the O/U-ratio. The effect of fission products previously thought to be “chemically inert” and thus not changing fuel chemistry has only recently been treated in modelling work [32]. This effect can safely be assumed to be significant for high burn-up UO, fuel, though the absolute values of changes in O/U-ratio are still being discussed and investigated. Two examples of radiation effects are given to further illustrate basic aspects of fuel performance: Fig. 9 shows in-pile, radiation enhanced diffusion of U and Pu. i.e. the rate-controlling species for matter transport in ceramic fuels: a temperature-independent athermal fission enhanced diffusion is observed below - 1000 o C in all ceramic fuels studied, if the experiments are performed in-pile. The measured D* values for UO, cannot be explained with conventional collision or thermal spike effects. However, a fast mobility of uranium interstitials in the pressure gradient of the spike away from its axis could help to explain the experimental data. The ratio of D * for UO, : UC : UN corresponds closely to that of their thermal conductivity, and is confirmed by in-pile creep data (e.g. ref. [33)). These measurements help thus to explain the differences in plasticity at the fuel rim (hence at the clad) during reactor operation between oxides, carbides and nitrides. Fig. 10 shows recent results [31] on the behaviour of radiation induced point defects and points out further interesting differences between (the largely ionic) UO, and (the largely metallic) UN. Point defect concentrations in radiation damaged (or quenched) specimens are usually deduced from changes in properties such as electrical resistivity or lattice parameter. Such measurements do not allow a unique attribution to which type of defect (e.g. U, 0 or N vacancies, etc.) is formed or dnnealed. Rutherford backscattering allows this distinction. Fig. 10 shows U defects only, in the form of peaks in the aligned spectra. The areas and widths of these peaks are measures for the number and location of displaced U atoms. Both UO, and UN were damaged
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance
127
Fig. 6. Lattice plane images in ion implanted UO, [24].
by ion implantation. In the case of UO, and high damage levels, only - 1 U defect per incoming ion survived the bombardment at room temperature; there-
neutral trivacancy
charged trivacancy
EJ
@EC,
possible sites for Xe
m ,L+
-, 9
I+
El q
0
U- vacancy 0 -vacancy Xe-atom
possible saddle point for Xe migration
Fig. 7. Equilibrium
lattice site of Xe in UO, migration path.
and
possible
fore, more than 99% of the defects formed recombined immediately at room temperature. The remaining defects did not leave the range profile of the bombarding ions, i.e. they moved at the most a few nm during implantation and at ambient temperature. The important instantaneous damage recovery is also the reason for the fact that lattice images such as those of fig. 6 can be made with ion implanted UO,, and that UO, as a nuclear fuel does not become amorphous (as many other ceramics do). In contrast to UO,, a fraction of the U defects are seen to migrate away from the damaged surface layer in UN, as is obvious from the “dechanneling knee” located at lower energies (smaller channel numbers) and thus deeper in the crystal than the damage peak (see upper scale in fig. 10). Such dechanneling knees are typical for metals for which is is known that defects are mobile below room temperature. For UN, the dechanneling knee shows that U defects (probably interstitials) are mobile at room temperature and migrate out of the ion-implanted damaged zone into the virgin crystal. Further channeling experiments at liquid N, temperature have shown that these U defects actually move below room temperature. In addition, as with
Hj. Matrke, H. Blunk / Fuel research and fuel in-pile perjormance
128
z-i10 g
Kr
-a
5.3~1dZ ,ons/cm’
.j
- -
ThOz
o
SIMFUEL
0.5 -
UO,, important instantaneous damage recovery occurs in UN as well to a large extent. These results explain that UN does not become amorphous either during irradiation and they also help explain the radiation-enhanced diffusion coefficients D* of fig. 9. To conclude this short list of examples, the direct measurement of radial oxygen profiles in LMFBR mixed oxide fuels and of the increase of the oxygen potential AG(0,) in such fuels is mentioned. A solid state galvanic micro-cell is used for these investigations. Very recently, the AG(0,) of (U, Pu)Oz fuel from the Phenix reactor with a very high burn-up of 11 at% was successfully measured [34]. The results indicated a linear increase of AG(0,) of the fuel (initial composition (U, sPuO z )O, 9X) up to 11 at% burn-up without saturation.
0 9
t a_p
e-6’
8’
“0’
100 I 5 9.10’1lals/crn*
i 5. Summary and concluding remarks
temperature, “C Fig. 8. Release of Rb. I and Kr from ion-implanted (AECL, 3 at% burn-up [2X].
--
2000 I
CA _E 2
SIMFUEL
1500 1300 1100 I I a
The approach of the “al] in-house” activity adopted at the European Institute for Transuranium Elements to understand the in-pile performance of nuclear fuels was described with a number of examples of experiments giving insight in basic aspects. Detailed careful characterization of as-produced fuel and of fuel irradiated to different burn-up with supporting laboratory and modelling work provided not only reliable data for
900 800 700 h I I I .A
GM 200 ,dr ,
radiatlon -10
-
thermally
activated
enhanced o”:
normalized
oxides
. UO,, *IU,,,
f/&s
. DJ,h,,lO,
IU,,,Pu,,lC,
o UC,
*
nitrides
0 UN
4 HJ,,,Pu,,)N
diffusion
oPV+
to ~=5~10’z
Py,4&,
carbides
I50 T,“C
Itl,,k~,,~E
* WY1 Wu,,,,P~,,,I UI
-20
Fig. 9. Radiation
enhanced
i
6
8 lo’-
ld”’ 21
23 lO?T,
“a
25 K-’
diffusion of U and Pu in the ceramic fuels UO,, (U, Pu)O,, UC, (U, Pu)C, UN, (U, Pu)N, U(C, N) and (U, Pu)(C, N). The data are normalized to a fission rate of 5 X 10” fissions/cm3 s.
H. Blank / Fuel research and fuel in-pile performance
Hj. Matrke,
a3
0.2 ,
1.4
0.1
0
depth below surface,pm 0.3 0.2 0.1 0
0.4
I
1.6
1.6
129
1:4 E,MeV
1:s
118
---I .
0 5110’~ ions/cm’
l
:I$
2~10”
ions/cm’
,,d
mi’ _ _cleaved
280
--_--___’
320
1
360
s,
1
4Ml
I
I
.---------
440
electropolished
390
43o
470 t channel number
Fig. 10. Rutherford backscattering spectra of UO, (left) and UN (right) single crystals implanted with high dose Xe-ions (s = surface peak due to surface atoms which are not shielded along atomic strings, d = damage peak due to displaced U atoms situated within the range of the implanted ions, d.k. = dechanneling knee due to migration of U-atoms into the undamaged crystal). Both scattering in a random direction and in a direction where a low index direction of the crystal is aligned with the He beam are shown (channeling mode).
basic physical and chemical properties but allowed also to define mechanisms and kinetics important for in-pile performance. Knowledge of these properties and of these mechanisms is indeed necessary to develop “tailor-made” fuels and to understand (and predict) the in-pile behaviour of nuclear fuels. A convincing example for fuel fabrication is the KWU process of oxidative sintering of UO,, which reduces sintering temperatures from 1700 to 1100 o C, reduces also sintering times, enables the use of cheap furnaces and sintering boats and saves energy as well, The process is based on fundamental measurements of U diffusion in UO*+, showing a largely enhanced U-mobility (and thus enhanced sintering) with increasing deviation x from stoichiometry [35]. The above knowledge is also of paramount importance for modelling work. Rather frequently, fuel performance codes use imcomplete, oversimplified or even physically incorrect models. This is often tolerable if interpolation is asked for within the range of data to which the model is fitted, but extrapolation beyond this range is likely to yield invalid results. In this sense, to have no model is better than to use an incorrect one. Also, sometimes
surprising new mechanisms are postulated based on single isolated observations without considering the existing knowledge on basic aspects of fuel in-pile performance. The example of postulated gas release at very low temperatures in transient-tested UO, (but based on too low calculated temperatures) has already been given above. (Many laboratory and hot cell anneals of UO, irradiated at low temperatures had shown that such unexpected releases do not occur.) Other examples are assumptions of high thermal solubilities of rare gases just because locally in the fuel absence of fission gas bubbles is observed, or the postulation of U,O, existing in high bum-up UO, (Any U,O, existing in UO, would be transformed into UO,+. at reactor operating temperatures and U,O, would not be stable during fission either). Equally, the seemingly straightforward “conclusion” that gas release from oxide fuel must be fast because the fuel is hyperstoichiometric UO,,, necessitates different pieces of basic work and knowledge before the implied relation can be accepted. It is true that Kr and Xe diffuse significantly faster in UO,,, than in UO, (27) but this observation was made for single gas atom diffusion at low gas concentrations. It is
130
HJ. Mairke,
H. Blank / Fuel research rend fuel in-prle performunce
also well known that gas release at higher burn-up is a complex process of those single gas atoms becoming trapped, precipitating into bubbles, being freed from the bubbles by radiation-enhanced re-solution, reaching grain boundaries after multiple precipitation-re-solution processes, or by sweeping or bubble diffusion, and then having to wait for grain boundary bubble venting by coalescence, tunnel formation, crack formation etc. Unless it is shown that all these processes are accelerated in UO,,,, or that single gas atom diffusion between traps and bubbles is rate-controlling for release (it is known that this is not generally the case for release from UO,), a simple comparison of rare gas diffusion coefficients will not be appropriate to predict release from hyperstoichiometric fuel. For lack of space, only a limited number of examples could be given to illustrate the current advanced state of understanding basic aspects of fuel in-pile performance. Most examples were for oxides, since these are the most frequently used fuels, and most examples were from work performed at the European Institute for Transuranium Elements since these proceedings feature its 25 th anniversary. The R&D approach of the Institute has helped to achieve a thorough understanding of the in-pile mechanisms of different nuclear fuels on e.g. densification and swelling, restructuring, release of fission gases and other volatile elements, thermodiffusion in the existing temperature gradient and radiation enhanced diffusion, mechanical and fracture behaviour, chemical behaviour of fission products etc. The results are used as in-put data in mechanistic fuel performance codes.
6. Outlook for future investigations The R&D approach described can easily be applied to promising future research on nuclear fuels. Particular areas that should be investigated in detail are (i) problems of spent nuclear fuel direct storage of LWR fuel (and possibly other existing fuel), (ii) for LWR’s: MOX fuel and high burn-up UO, fuel, (iii) for LMFBRs: nitride fuels, and (iv) for nuclear waste aspects: questions related to transmutation of higher actinides (minor actinides) in suitable nuclear fuels.
Acknowledgements The authors would like to thank the numerous colleagues of the Institute who contributed with experimental and theoretical work to increase the knowledge
on basic aspects of fuel in-pile performance. The different contributing laboratories and the responsible scientists are listed in the introduction of this article. Thanks are also due to a large number of scientists from other laboratories (KfK, Germany; CEA, France: AERE. Harwell; AECL, Canada; ANL, USA and others, as well as those mentioned in the list of references).
References [I] Hj. Matzke. H. Blank, M. Coquerelle,
K. Lassmann, I.L.F. Ray, C. Ronchi and C.T. Walker, in: these Proceedings (TM Symposium), J. Nucl. Mater. 166 (1989) 162. [2] K. Lassmann, C. Ron&i and G. Small, in: these Proceedings (TTT Symposium). J. Nucl. Mater. 166 (1989) 109. [3] H. Blank. K. Richter, M. Coquerelle, Hi. Matzke, M. Campana, C. Sari and I.L.F. Ray, in: these Proceedings (TlT Symposium), J. Nucl. Mater. 166 (1989) 92. G. Kramer, C. Sari and P. [41 K. Richter, J. Gueugnon, Werner, Nucl. Technol. 70 (1985) 4. 151 C. Sari, Euratom Report EUR-5812 EN (1978). [61 P. Werner and H. Blank, Nucl. Technol. 52 (1981) 73. and P. [71 H. Blank, R. Mandler, Hj. Matzke, J.L. Routbort Werner. Res. Mechanica 5 (1983) 137. PI Hj. Matzke, V. Meyritz and J.L. Routbort, J. Am. Ceram. Sot. 66 (1983) 183. [91 Hj. Matzke. J. Nucl. Mater. 114 (1983) 121. and Hj. Matzke, Solid State Ionics 12 [lOI D. Glasser-Leme (1984) 217. 1111 Hj. Matzke, Radiat. Eff. 75 (1983) 317. in Non-Stoichiometric Comu21 Hj. Matzke, in: Transport pounds, Eds. G. Simkovich and V.S. Stubican (Plenum, New York, 1984) p. 331. v31 R.W. Ohse. J.F. Babelot. C. Cercignani, J.P. Hiernaut, M. Hoch, G.J. Hyland and J. Magill, J. Nucl. Mater. 130 (1985) 165. u41 J.F. Babelot, R.W. Ohse and M. Hoch, J. Nucl. Mater. 137 (1986) 144. u51 C.T. Walker and K. Lassmann, J. Nucl. Mater. 138 (1986) 155. Science of Advanced LMFBR Fuels, A [I61 Hj. Matzke, Monograph on Solid State Physics, Chemistry and Technology of Carbides, Nitrides and Carbonitrides of Uranium and Plutonium (North Holland, Amsterdam, 1986). u71 C. Ronchi and C. Sari, J. Nucl. Mater. 50 (1974) 91. [181 C. Ronchi and Hj. Matzke, in: Fuel and Fuel Elements for Fast Reactors, Vol. I (IAEA, Vienna, 1974) p. 57. v91 Hj. Matzke and C. Ronchi, Europ. Appl. Res. Rep. 5 (1984) 1105. PI R.A. Verrall, Hj. Matzke, T. Ogawa and B.J.F. Palmer, Canada Report AECL-9475 (1986) p. 558. [21] Hj. Matzke, I.L.F. Ray and R.A. Verrall, Report Int. Atom Energy Agency, IWGFPT/27 (IAEA, Vienna. 1987) p. 183.
Hj. Matzke, H. Blank / Fuel research and fuel in-pile performance [22] I.L.F. Ray, H. Thiele and Hj. Matzke, unpublished results. [23] M. Mogensen and C.T. Walker, personal communication, presented at Enlarged Halden Meeting, Loen, May 1988. [24] Hj. Mat&e, C. Rossouw and S.E. Donnelly, unpublished results. [25] Hj. Matzke, Radiat. Eff. 53 (1980) 219. [26] R.A. Jackson and C.R.A. Catlow, J. Nucl. Mater. 127 (1985) 161, 167. [27] R. Lindner and Hj. Matzke, Z. Naturforschung 14a (1959) 582; AERE-Trans 845 and CEA-tr-A-856 (1959). 1281 Hj. Matzke, R.A. VerraII and B.J.F. Palmer, unpublished results.
[29] [30] [31] [32]
131
I. Misfeldt, J. Nucl. Mater. 135 (1985) 1. M. Coquerelle, personal communication. Hj. Matzke, S. Fritz and A. Turos, unpublished results. D.A. Mac Innes and P.W. Winter, J. Phys. Chem. Solids 49 (1988) 143. [33] D. Brucklacher and W. Dienst, J. Nucl. Mater. 42 (1972) 285. [34] Hj. Matzke, J. Ottaviani, D. Pellottiero an J. Rouault, J. Nucl. Mater. 160 (1988) 142. [35] Hj. Matzke, Euratom Report EUR-7700 DE (1982) p. 139.