CONCLUSION AND OUTLOOK - SYNTHESIS OF THE RESULTS OBTAINED ON THE ADVANCED UO 2

Stress, MPa

SYNTHESIS OF THE RESULTS OBTAINED ON THE ADVANCED UO 2

5. CONCLUSION AND OUTLOOK

Different kinds of undoped and doped UO2 fuels were fabricated and irradiated in the TANOX device, in the Siloé reactor. The temperature at the centre of the annular pellets was maintained under 700°C until the end of the irradiation so as to avoid activation of thermal gas diffusion and release. The pellets, which had reached a burn-up of about 10 GWd/tU, were then submitted to post-irradiation annealing tests at 1700°C intended to stimulate fission gas release and to test their retention capacity. The ⁸⁵Kr release was measured by gamma spectrometry after 30 minutes and 5 hours at 1700°C, and the samples were examined by optical micrography.

The results show that the fuels with best fission gas retention are characterized by both a large grain size (over 50 µm) and intragranular sites which are favourable to gas bubble nucleation and pinning. These characteristics are achieved in (i) large grained UO2.01and UO2.02fuels in which structural defects due to the hyperstoichiometry are favourable to bubble nucleation, (ii) UO2 doped with 0.2% Cr2O3 which acts as a grain growth activator and also forms second phase precipitates. Gas release can even be lower if intergranular gas retention is improved by the addition of a small quantity of silica.

On the basis of these results, several microstructures were selected for a further study of gas retention at high burn-up: UO2 doped with 0.2% Cr2O3, submitted or not to a post-sintering reducing anneal, UO2 doped with 0.2% Cr2O3 and 0.02 % SiO2 and UO2.02 doped with 0.2 % Cr2O3. These fuels were irradiated in the TANOXOS device located in the Osiris reactor, to a burn-up of 65 GWd/tU. After irradiation, fuel samples were submitted to annealing tests; the first recent results show that doped stoichiometric fuels have a better gas retention capacity than undoped fuel and prove their interest for high burn-up conditions.

Besides these analytical studies, an irradiation of doped UO2 fuels was launched in order to test their behaviour in a PWR (CONCERTO program, directed by FRAMATOME-ANP in cooperation with EDF and the CEA). The fuels were doped with Cr2O3, SiO2, and/or Al2O3

and they were manufactured by FBFC.

Some power ramp tests were performed on chromium oxide doped UO2 segments, which had been discharged after two PWR irradiation cycles [17]. These ramps aimed at testing the behaviour of the fuel in conditions of pellet-cladding mechanical interaction. The segments could withstand very high ramp power levels without failure. Moreover, rod punctures performed after ramp tests showed a lower fission gas release compared to standard UO2, which confirms the improved fission gas retention capacity of the doped fuel.

These very promising results concerning the PCMI performance and the improved fission gas retention capacity of chromia doped fuel led FRAMATOME-ANP to launch in 2001 the irradiation of lead fuel rods containing chromium oxide fuel pellets loaded in M5^TM cladding tubes. This irradiation aims at obtaining all the data required to demonstrate the ability of this advanced fuel to become a high burn-up and PCMI remedy.

ACKNOWLEDGEMENTS

The authors would like to acknowledge all the persons involved in the fabrication, irradiation, and examinations of the advanced UO2 fuels. We especially wish to thank S. Ravel for the annealing tests and G. Eminet for the micrographic examinations.

REFERENCES

[1^] Ph. DEHAUDT, L. CAILLOT, G. DELETTE, G. EMINET, A. MOCELLIN,

“Irradiation of UO2+x fuels in the TANOX device”, Advances in Pellet Technology for Improved Performance at High Burnup, Proc. IAEA Tech.l Com. Mtg Tokyo, Japan, 28 Oct–1 Nov. 1996), IAEA-TECDOC-1036, Vienna (1998) 277.

[2] L. BOURGEOIS, Contribution à l’étude du rôle des dopants dans la densification et la croissance cristalline du dioxyde d’uranium (Contribution to the Investigation of Densidication of Crystalline Uraniume Dioxide by Dopants), Thesis, Institut National Polytechnique de Grenoble, 1992.

[3] L. BOURGEOIS, PH. DEHAUDT, C. LEMAIGNAN, A. HAMMOU, Factors governing microstructure development of Cr2O3-doped UO2 during sintering, Journal of Nuclear Materials, 297 (2001) 313–326.

[4] V. PERES, Contribution à l’étude de la dispersion de particules de phases secondaires dans le dioxyde d’uranium polycristallin (Contribution to the Investigation of 2^nd Phase Particles in Polycrustalline Uraniume Dioxide), Thesis, Institut National Polytechnique de Grenoble, 1994.

[5] V. PERES, L. BOURGEOIS, PH. DEHAUDT, Grain growth and Ostwald ripening in chromia-doped uranium dioxide, Journal de Physique IV, colloque C7, vol. 3 (nov.

1993).

[6] C. DUGUAY, A. MOCELLIN, PH. DEHAUDT, M. SLADKOFF, High temperature mechanical tests performed on doped fuels, Proc. IAEA Tech.l Com. Mtg Tokyo, Japan, 28 Oct–1 Nov. 1996), IAEA-TECDOC-1036, Vienna (1998) 409.

[7] C. VIVANT-DUGUAY, Contribution à l’étude du fluage du dioxyde d’uranium. Rôle des activateurs de croissance cristalline, (Contribution to the Investigation of Creep of Uraniume Dioxide, Role of Activators )Thesis, Institut National des Sciences Appliquées de Lyon, 1998.

[8] P.T. SAWBRIDGE, C. BAKER, R.M. CORNELL, K.W. JONES, D. REED, J.B.

AINSCOUGH, The irradiation performance of magnesia-doped UO2 fuel, Journal of Nuclear Materials, 95 (1980) 119–128.

[9] R. YUDA, H. HARADA, M. HIRAI, T. HOSOKAWA, K. UNE, S. KASHIBE, S.

SHIMIZU, T. KUBO, Effects of pellet microstructure on irradiation behavior of UO₂ fuel, Journal of Nuclear Materials, 248 (1997) 262–267.

[10] K. UNE, S. KASHIBE, K. ITO, Fission gas behavior during postirradiation annealing of large grained UO2 fuels irradiated to 23 GWd/t, Journal of Nuclear Science and Technology, 30 [3] (1993) 221–231.

[11] J.C. KILLEEN, Fission gas release and swelling in UO2 doped with Cr2O3, Journal of Nuclear Materials, 88 (1980) 177–184.

[12] S. KASHIBE, K. UNE, Effect of additives (Cr2O3, Al2O3, SiO2, MgO) on diffusional release of ¹³³Xe from UO2 fuels, Journal of Nuclear Materials, 254 (1998) 234–242.

[13] K. UNE, I. TANABE, M. OGUMA, Effects of additives and the oxygen potential on the fission gas diffusion in UO2fuel, Journal of Nuclear Materials, 150 (1987) 93–99.

[14] J.H. DAVIES, E.V. HOSHI, D.L. ZIMMERMAN, Ramp test behavior of high O/U fuel, Journal of Nuclear Materials, 270 (1999) 87–95.

[15] S. VALIN, Etude des mécanismes microstructuraux liés au relâchement des gaz de fission du dioxyde d’uranium irradié, (Investigation of Mechanism of FGR from Irradiated Uraniume Dioxide) Thesis, Institut National Polytechnique de Grenoble, 1999.

[16] S. VALIN, A. MOCELLIN, G. EMINET, S. RAVEL, J.C. JOUBERT, Modelling the behaviour of intergranular fission gas during out-of-pile annealing, Fission Gas Behaviour in Water Reactor Fuels (Proc. OECD/NEA-IAEA Seminar Cadarache, France, 26–29 Sept. 2000), OECD/NEA, Paris (2001).

[17] Ch. DELAFOY, P. BLANPAIN, C. MAURY, PH. DEHAUDT, CH. NONON, S.

VALIN, Advanced UO2 fuel with improved PCI resistance and fission gas retention capability, TOPFUEL 2003, (Proc. Int. Conf. Würzburg, Germany, 16-19 March 2003), Inforum GmbH, Germany, CD-ROM.

[1] Ph. DEHAUDT, L. CAILLOT, G. DELETTE, G. EMINET, A. MOCELLIN, “Irradiation of UO2+x fuels in the TANOX device”, Advances in Pellet Technology for Improved Performance at High Burnup, Proc. IAEA Tech.l Com. Mtg Tokyo, Japan, 28 Oct–1 Nov. 1996), IAEA-TECDOC-1036, Vienna (1998) 277.

[3] L. BOURGEOIS, PH. DEHAUDT, C. LEMAIGNAN, A. HAMMOU, Factors governing microstructure development of Cr2O3-doped UO2 during sintering, Journal of Nuclear Materials, 297 (2001) 313–326.

[5] V. PERES, L. BOURGEOIS, PH. DEHAUDT, Grain growth and Ostwald ripening in chromia-doped uranium dioxide, Journal de Physique IV, colloque C7, vol. 3 (nov. 1993).

[8] P.T. SAWBRIDGE, C. BAKER, R.M. CORNELL, K.W. JONES, D. REED, J.B. AINSCOUGH, The irradiation performance of magnesia-doped UO2 fuel, Journal of Nuclear Materials, 95 (1980) 119–128.

[9] R. YUDA, H. HARADA, M. HIRAI, T. HOSOKAWA, K. UNE, S. KASHIBE, S. SHIMIZU, T. KUBO, Effects of pellet microstructure on irradiation behavior of UO2 fuel, Journal of Nuclear Materials, 248 (1997) 262–267.

[11] J.C. KILLEEN, Fission gas release and swelling in UO2 doped with Cr2O3, Journal of Nuclear Materials, 88 (1980) 177–184.

[12] S. KASHIBE, K. UNE, Effect of additives (Cr2O3, Al2O3, SiO2, MgO) on diffusional release of ¹³³Xe from UO2 fuels, Journal of Nuclear Materials, 254 (1998) 234–242.

[13] K. UNE, I. TANABE, M. OGUMA, Effects of additives and the oxygen potential on the fission gas diffusion in UO2fuel, Journal of Nuclear Materials, 150 (1987) 93–99.

[14] J.H. DAVIES, E.V. HOSHI, D.L. ZIMMERMAN, Ramp test behavior of high O/U fuel, Journal of Nuclear Materials, 270 (1999) 87–95.

[17] Ch. DELAFOY, P. BLANPAIN, C. MAURY, PH. DEHAUDT, CH. NONON, S. VALIN, Advanced UO2 fuel with improved PCI resistance and fission gas retention capability, TOPFUEL 2003, (Proc. Int. Conf. Würzburg, Germany, 16-19 March 2003), Inforum GmbH, Germany, CD-ROM.

FISSION GAS RELEASE FROM HIGH BURN-UP UO₂ FUELS UNDER SIMULATED OUT-OF-PILE LOCA CONDITIONS

Y. PONTILLON, D. PARRAT, M.P. FERROUD PLATTET S. RAVEL¹, G. DUCROS, C. STRUZIK

CEA Lez Cadarache, Saint Paul Durance A. HARRER EDF/SEPTEN/TE Villeurbanne Cedex, France

Abstract

A specific emphasis has been put recently in France on mechanisms, which promote Fission Gas Release (FGR) in Loss Of Coolant Accident (LOCA) thermal conditions. One of the most useful information for understanding the FGR mechanisms is to achieve reliable tests to measure the absolute level and the time dependence of the released gases and the corresponding fuel micro-structural changes during representative thermal transients. In order to obtain these data, the so-called

“GASPARD” programme has started at the beginning of 2000 at the CEA, in collaboration with EDF.

The first step of this programme deals with short sections of UO₂ fuel rods irradiated in power reactors during 4 and 6 cycles. The experimental process consists in annealing the fuel samples in out-of-pile conditions, from 300°C up to a temperature ranged between 900°C and 1600°C, with a wide range of temperature increase rates (from 0.2°C/s up to 20°C/s). The upper temperature is maintained during several minutes or is immediately decreased. The release rate of ⁸⁵Kr is measured by on-line gamma spectrometry and delayed gamma spectrometry on a gas sample. Stable isotopes are measured by gas chromatography coupled to a mass spectrometry. Specific post-test observation devices are then used (e.g. metallographic inspection, Scanning Electron Microscopy). The present communication focuses firstly on the description of the GASPARD experimental process, then on experimental facility implemented at the CEA/DEN for fast transient and LOCA reproductions. After that, the results regarding 4 cycles UO₂ fuel are presented and discussed, and a good agreement with the CEA code METEOR calculations is pointed out.

1. INTRODUCTION

1.1. R and D requirements on fuels for current reactors

The current generation of LWRs have operated very successfully for over 30 years, and with plant lifetime extension programmes (PLEX), individual reactors are expected to reach a lifetime approaching 60 years, i.e., up to the year 2030. Regarding the fuel properties and the fuel behaviour under neutron flux, much of the start-up R&D has been completed, although safety issues appropriate to high burn-up are still being addressed.

However, on the short and medium-term, the liberalisation of the electricity market lead operators and fuel vendors to consider a variety of means to enhance plant performance and reduce costs. This trend has taken place in the last several years and is likely to continue in the long term. In the fuel area it is supported by a few basis orientations, such as burn-up extension of standard fuels, new types of fuel, more efficient fuel management schemes, new cladding materials and new fuel assembly designs.

1 Laue Langevin Institute (ILL) –Reactor Division - 6 rue J. Horowitz – BP 156 - F-38042 Grenoble Cedex 9.

As a consequence of this evolution, there is a permanent need to reassess the reactor safety studies, in order to ensure that these progresses take place without detriment to safety. This includes improving the knowledge and upgrading the calculation tools used for safety assessment. Utilities will need to continuously review and to update eventually the technical basis and the assumptions used for the design analysis of the fuel core. This will require not only that the database for transient analyses is extended to broader conditions and materials, but also that the uncertainties associated with such database and its use for specific applications is adequately quantified. A successful strategy to reach these objectives is to gain reliable experimental data from separate effects experiments carried out in Material Test Reactors (MTRs) or in hot cell laboratories.

1.2. Importance of studies on Fission Gas Release 1.2.1. High burn-up UO₂ fuel behaviour

The fuel enrichment limit in force today is equivalent to less than 5 wt% ²³⁵U. This limits fuel assembly discharge burn-up to around 60 to 65 GWd.t_U^-1. Most commercial reactors have a lower burn-up limit, and support work is in hand to improve effectiveness of the standard fuel and to approach the limit set by the enrichment. Several evolved products are in testing in Material Test Reactors (MTRs) and in power reactors:

(i) improved UO₂, with large grain size including or not additives in the matrix or at the grain boundaries,

(ii) integrated Burnable Absorber Fuels, such as uranium-gadolinium fuels.

There are two principal concerns: rod over-pressure due to the high inventory of fission gases at high burn-up and the potential behaviour of the High Burn-up Structure (HBS). The latter is first observed in the pellet rim at pellet averaged burn-up levels ~ 50 GWd.t_U^-1 and extends radially inwards at higher burn-up. The concern for normal operation is that it may increase slightly fuel temperatures and also fission gas release by acting as a thermal barrier due to the poor thermal conductivity of the HBS.

1.2.2. Introduction of MOX

A prerequisite for the loading of MOX fuel in current reactors is that MOX safety requirements should not be different from those required for UO₂. Studies to date have shown that there is little difference in the behaviour of the two types of fuel. However, whereas MOX is more compliant, resulting in reduced PCMI, it has a higher remnant reactivity at high burn-up, hence powers, fuel temperatures and fission gas release (FGR) are likely to be greater than for UO₂ fuel in the event of a transient.

1.2.3. The FGR GASPARD programme at CEA

Fission gas release is an important input data of the fuel licensing process, either in normal operation or as the radioactive “source term”, relatively to the consequences of a nuclear incident and/or accident on the surrounding populations as well as on the environment. For this purpose, several R&D programs have been recently initiated in France through joint actions between the Commissariat à l’Energie Atomique (CEA) and Electricité de France (EDF), including occasionally a support from Framatome-ANP or the Nuclear Radioprotection and Safety Institute (IRSN).

Within this framework, a specific emphasis has been put recently on mechanisms, which promote the Fission Gas Release (FGR) in Loss Of Coolant Accident (LOCA) thermal conditions, mainly thanks to the so-called “GASPARD” programme. One of the most useful information for understanding the FGR mechanisms is to achieve reliable tests to measure the absolute level and the time dependence of the released gases and the corresponding fuel micro-structural changes during representative thermal transients. This database is used:

- to define the fission product (FP) “source term” out of the damaged rods during a given type of accidental sequence,

- to verify if existing safety criteria are well adapted to new fuels,

- to enhance models predicting behaviour of various fuels during an accidental sequence, thanks to the understanding and quantifying of the basis mechanisms.

2. DESCRIPTION OF THE GASPARD EXPERIMENTAL PROCESS

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