Abstract. One of the most important areas of research concerning severe accidents in PWR is certainly quantifying the source term of radioactive materials. This is in great part due to the consequences of the Three Mile Island (1979), Chernobyl (1986) and more recently Fukushima (2011) accidents.

A large number of research programs have thus been undertaken on this subject with particular efforts devoted to understanding the mechanisms that lead to the release of fission products (FP). In France, the HEVA-VERCORS programs, funded by IRSN and EdF, were initiated by the CEA and conducted up to 2002.

However, major uncertainties still remained. It has thus been decided to start a new co-operative research program, called "International Source Term Program (ISTP)", based on separate-effect experiments. Regarding the source term quantification, a new experimental device, VERDON, has been built in the LECA-STAR facility (CEA Cadarache site) and four VERDON tests are considered: Three of them are dedicated to the FP release from high burn-up UO 2 fuel and from MOX fuel, one of them is dedicated to the impact of air ingress on the FP release from the fuel and their transport into the primary circuit.

The present paper describes the first VERDON test performed at the end of September 2011 with special emphasis on FP and actinides behavior during the accidental sequence itself (a second paper deals details the results regarding post-test examination (SEM, EPMA and SIMS) of the VERDON-1 sample). This test, devoted to UO 2 fuel behavior and fission product releases under reducing conditions at very high temperature (~2610°C), has validated the very good performances of the VERDON loop. It did not result in a relocation of the fuel sample at the end of the test. According to the FP behavior measured by the on-line gamma station (fuel sight), the FP general classification, in relation to their released fraction, is very well highlighted together with the burn up effect on the release rate.

Introduction

One of the most important areas of research concerning a severe accident in a pressurized water reactor (PWR), on both a French and an international level, is determining the source term, i.e.

quantifying the nature, release rate and global released fraction of fission products (FP) and other radioactive materials. This is in great part due to the consequences of the Three Mile Island (1979), Chernobyl (1986) and more recently Fukushima accidents. In the course of this type of scenario, the chain of events can effectively result in primary coolant boiling and draining, meaning that it is no longer cooling the core. One direct result is the core melting and the release of FP and structural and/or activated control rod elements (e.g. activation products, AP) into the containment building. If there is a failure in the various protective barriers, the FP/AP can leak out of the containment building and be released into the environment.

A large number of research programs have thus been undertaken on this subject in various countries. In line with this approach, IRSN (France) has been the driving force and has conducted programs specifically focusing on determining the source term, with particular efforts devoted to understanding the mechanisms that lead to the release of FP, since only a very complete knowledge of the phenomena governing the behavior of FP/AP under such constraints will make it possible to define the actions that need to be planned (and/or performed) to minimize emissions and optimize the protection of both the people and the environment. The HEVA [START_REF] Leveque | The HEVA experimental programme[END_REF]/VERCORS [2] programs were thus initiated by the CEA. VERCORS has considerably broadened the field of application by exploring higher temperatures and by testing a wider range of fuels (UO 2 , MOX, debris bed configurations, high burn-ups) in a more complex experimental installation with better instrumentation. It was composed of 17 tests which were conducted over 14 years, in accordance with 3 experimental phases. A first series of six tests (VERCORS 1 to VERCORS 6, Table 1) conducted between 1989 and 1994 on UO 2 fuel in a higher temperature range than the previous HEVA phase, close to the fuel relocation [START_REF] Ducros | Fission Product release under severe accidental conditions; general presentation of the program and synthesis of VERCORS 1 to 6 results[END_REF]. This series made it possible to integrate certain FP with low volatility into the HEVA results database. Two series of tests -VERCORS HT and RT (Table 2)conducted alternately throughout 1996-2002 allowing the data base extension up to the less volatile FP.

These analytic experiments simulating severe PWR accidents were financed jointly by IRSN and EDF. Their aim was to quantify the released fraction and release rates of FP from irradiated nuclear ceramics (UO 2 or MOX, typically three PWR pellets in their original cladding), determine the nature of the gases and aerosols emitted (particle size analysis and speciation), and understand the fuel degradation mechanism. These experimental sequences were carried out in a very high activity cell and were commonly considered to be complementary to the PHEBUS FP [START_REF] Schwarz | PHEBUS FP: a severe accident research programme for current and advanced light water reactors[END_REF] integral tests and comparable with certain tests carried out abroad: HI/VI [START_REF] Lorenz | A summary of ORNL fission product release tests with recommended release rates and diffusion coefficients[END_REF] in the United States, VEGA [START_REF] Kudo | VEGA; an experimental study of radionuclides release from fuel under severe accident conditions[END_REF] in Japan or the program conducted in Canada [START_REF] Lui | A summary of CRL fission product release measurements from UO2 samples during post-irradiation annealing[END_REF]. The experimental results of this program are used to (a) define the envelope values for released fraction within the scope of assessing reference source terms for all French PWRs, and (b) validate the semi-empirical or mechanistic models regarding FP release and transport while qualifying the simulation codes by integrating these models [START_REF] Veshchunov | Development of the mechanistic code MFPR for modelling fission product release from irradiated UO2 fuel (part 1). Development and validation of new models[END_REF], [START_REF] Dubourg | Development of the mechanistic code MFPR formodelling fission product release from irradiated UO2 fuel (part 2). Application to integral tests VERCORS 4-5 and PHEBUS FPT0[END_REF], [START_REF] Brillant | Ruhtenium release from fuel in accident conditions[END_REF], [START_REF] Beuzet | Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code[END_REF].

However, major uncertainties still remained in some fields, concerning the assessment of risks for populations and the environment [START_REF] Clement | The Phebus Fission Product and Source Term International Programs[END_REF]. As a consequence, it has been decided to As far as the source term quantification is concerned, four VERDON tests were considered. They are devoted to FP release from high burn up UO 2 fuel, MOX fuels and air ingress scenario. They will be performed in the new VERDON laboratory at the CEA Cadarache center.

The present paper deals with the VERDON-1 test itself. The main issue addressed by this first test concerns high burn up UO 2 fuel behavior -and corresponding fission product releases -under reducing conditions at very high temperature (up to 2610°C). Moreover, the first part of the test (i.e. up to the end of the so-called oxidation plateau at 1500°C) is performed under the same atmosphere conditions compared to VERCORS RT6 test, which was conducted with a very similar UO 2 high burn up fuel, in order to check the continuity between VERCORS and future VERDON data bases [START_REF] Leveque | The HEVA experimental programme[END_REF]- [START_REF] Ducros | Fission Product release under severe accidental conditions; general presentation of the program and synthesis of VERCORS 1 to 6 results[END_REF]. The second and third parts of this article [START_REF] Schwarz | PHEBUS FP: a severe accident research programme for current and advanced light water reactors[END_REF], [START_REF] Lorenz | A summary of ORNL fission product release tests with recommended release rates and diffusion coefficients[END_REF] deal in details with the results regarding post-test examination (SEM, EPMA and SIMS) of the VERDON-1 sample. The second part focuses on the fuel behavior during the VERDON-1 test and the third part describes a promising methodology to assess non γ-emitter FP release thanks to post-test characterizations.

In the first part of this paper, the experimental setup is described (section 2). In the second part (section 3), both the fuel sample and the progress of the accidental sequence are presented.

In section 4, the results regarding FP release and fuel behavior are highlighted. The main results are discussed in the last part of the paper with a special focus on: (1) fuel relocation, (2) continuity between VERCORS and VERDON and (3) FP volatility and release kinetics.

VERDON Experimental setup

In the present part, the VERDON experimental circuit and the apparatus used to measure FP release / behavior are detailed successively.

Experimental circuit

The VERDON laboratory, implemented at the LECA-STAR facility, is constituted of 2 high activity cells (called C4 and C5) and a gloves-box, as illustrated in Figure 1. The C4 cell is dedicated to the sample reception, pre/post tests gamma scanning and loop elements storage.

The C5 cell contains the experimental circuit itself (i.e. VERDON Loop). It is dedicated to the accidental sequence realization and to on-line measurements. The glove-box main functions are to analyze and store the fission and carrier gases. The VERDON Loop in its release configuration (as used for VERDON 1 test) is illustrated in Figure 2. This experimental loop is constituted of (along the path of gas flow): (1) The fluid injection system, (2) The furnace (see below), (3) An aerosol filter located directly on the top of the furnace. Its filtering part is constituted of stainless steel poral  which function is to stop all the fission products under aerosol form. The aerosol filter is heated at 175°C10%, (4) A May-Pack filter, half part of this filter is filled with zeolite (impregnated with silver) in order to stop potential molecular iodine, the other part is empty and has been used as a gas gamma spectrometry sighting, even though the design and detector are not well suited; the May-Pack is heated at 150°C10% to avoid condensation, (5) A condenser whose function is to condense steam from the experimental gas and to recover the water for analysis, (6) A final safety filter which filtering part is constituted of stainless steel poral  , in order to stop any residual trace of fission products (other than gaseous Xe, Kr), [START_REF] Lui | A summary of CRL fission product release measurements from UO2 samples during post-irradiation annealing[END_REF] Upstream from the condenser, the circuit is constituted of stainless steel tubes heated at 150°C10%, (8) Downstream, the condenser is linked to the final safety filter thanks to a flexible stainless tube at ambient temperature, (9) Outside of the cell, a "linking line" is used to make the junction between the C5 cell and the gloves box. This gloves box is equipped with a flow-meter measuring the total flow-rate of the loop, a pressure sensor and a safety filter similar to the final safety filter of C5 cell (at the end of the release experimental loop). It is mainly dedicated to gas analysis and storage.

The VERDON furnace is based, as the previous VERCORS one, on induction technology [START_REF] Leveque | The HEVA experimental programme[END_REF].

Schematically (Figure 3), it is constituted of a coil surrounding a tungsten susceptor tube, which is the heating component of the furnace. A high frequency power supply generates a current in the coil. By electro-magnetic coupling, a current is generated into the susceptor tube and the corresponding electric energy is converted to thermal energy by Joule effect, leading to the heating of the susceptor tube. Then, the fuel sample, located inside the susceptor tube, is heated, mostly by thermal radiation.

The external part of the furnace is constituted of a quartz tube sealed between 2 stainless steel bases thanks to 2 joints1 . The internal part can be schematically divided into two areas delimited by a ceramic column. This latter is constituted by a double stack of concentric dense ceramics; in the central area, where the furnace temperature is the highest, the ceramics are in hafnium dioxide and at the ends, where the furnace temperature is the lower, they are in zirconium dioxide. The crucible, in which the fuel sample is introduced before the sequence, is located inside this ceramic column together with several cups allowing crucible's supporting and gases flow circulation. The susceptor and a double stack of concentric porous zirconia and hafnia are located outside of the ceramic column and are devoted to furnace insulation. A "susceptor gas" flow constituted of He circulates between the quartz tube and the "ceramic column" through the porous insulators and along both sides of the susceptor tube. It is maintained in slight overpressure regarding the experimental gas in order to create a protection of the susceptor tube (in tungsten) against oxidation from the experimental fluid.

In order to evaluate the fuel sample (center) temperature, a pyrometer, sighting under the crucible and 3 thermocouples (TC) located inside the insulators of the furnace are used. Two of these TC are in front of the fuel sample and the 3 rd is underneath (and used as a safety TC, in case the 1 st two ones were unusable). These TC are used to monitor the power supply of the furnace at low temperature, up to 1000°C. At higher temperature, the TC give indicative values which can be used to check the consistency of the pyrometer measurement, since the pyrometer is used to monitor the power supply of the furnace at high temperature, from 1000°C up to 2610°C.

Fission Product measurements

In the case of VERDON-1 configuration, FP release kinetics was measured by means of three complementary on line gamma spectrometry stations and one micro gas chromatography apparatus:

1. One gamma station was aimed directly at the fuel sample and used during the entire test. This gamma station makes it possible to measure the FP remaining in the fuel as a function of temperature (presented in section 4.1), which explains why a relatively imprecise quantification of the release kinetics was obtained 2 . The two advantages of this station come (i) from its ability to measure directly at the source (all the FP were measured, unlike at the other stations where deposits upstream could occur) and (ii) for its ability to indicate the precise moment when the fuel relocates by detecting the disappearance (or significant decrease) in the signal from non-volatile FP. This last point was well illustrated in the case of the VERCORS series [2,[START_REF] Pontillon | Study of the active role played byUO2-ZrO2-PF interactions on irradiated fuel collapse temperature from VERCORS tests[END_REF]. This gamma station is also used in order to perform pre and post qualitative gamma scanning of the sample respectively before and after the accidental sequence (presented in section 4.2).

2. One other gamma station was aimed at the large-capacity aerosol filter. It provides a very precise measurement of the FP deposited at this point, where most of the volatile FP were found (presented in section 4.1). It is highly complementary with the previous station (fuel sight).

3. The last gamma station was aimed at the may-pack. The may-pack design has been performed in order to allow measuring potential iodine deposits (after test).

4. The gas analysis can be also performed on-line by a micro gas chromatograph (µGC) (or sequentially by 4 sampling aliquots) localized inside the gloves box. The µ-GC extends the analysis of active gases to all the gases. Within the context of VERDON programs, the µ-GC can analyze H 2 (for concentration less than 1%), Kr and Xe.

In addition to the on-line gamma spectrometry measurements described above, which give access to FP release kinetics, the overall released fractions are obtained by means of quantitative gamma spectrometry of all the loop components. The measurements are carried out on a gamma scanning bench located in the C4 high activity cell of the VERDON laboratory. The initial inventory (see section 3.1 for more details) of the FP was first of all established by scanning the fuel sample before the experimental sequence 3 and was completed by calculations for the nongamma-emitting elements. Calculating the production of FP also allows (i) validating the coherence of the FP measurements and (ii) recalculating the irradiation conditions in PWR (fuel burn-up) and in the MTR (re-irradiation power level).

Fuel sample characteristics and progress of the accidental sequence

VERDON-1 fuel sample characteristics

The sample used was taken from UO 2 fuel irradiated up to around 72 GWd.t HM -1 (6 cycles) in a PWR operated by EDF. The UO 2 fuel pellets were manufactured by standard industrial process. The 235 U initial enrichment was 4.5 %. The sample was made up of a fuel rod section taken at the span 4 of the initial rod (FX0GAC-E04, irradiated 6 cycles in EdF Gravelines-5, Figure 4). It consists of two irradiated pellets in their original cladding (M5 alloy). Two half-pellets of depleted (and un-irradiated) uranium oxide are placed at each end of the sample and held there by crimping the cladding (Figure 4). Thus the cladding is not fully sealed. At this stage, it is important to note that the sample used in this test is very similar to the one involved in the VERCORS RT6 test (rod FX0GAC-N05 i.e. same fuel assembly, same power history, very closer burn up, Table 2).

Before the experimental sequence, the sample has been re-irradiated at low linear power in the OSIRIS material testing reactor for ten days, in order to recreate the short half-life FPs without any in-pile release. As a consequence, these FPs (i.e. 99 Mo, 132 Te, 133 I, 131 I, 140 Ba…), important for their radiobiological effects, are measurable by using on-line gamma spectrometry during the experiment.

The initial FP inventory, which is used as reference to calculate the released fraction, is determined in two different ways, depending on the evaluated FP. The first way is to measure, by quantitative gamma spectrometry the contents of the VERDON-1 sample (before and after the re-irradiation in MTR). The second way is by calculating the FP content of the sample by a so-called evolution code like CESAR [START_REF]DEN/DER/SPRC[END_REF]. This was done because first, some fission products cannot be measured inside the pellet, 85 Kr for instance as well as all stable FP and/or pure  emitters, but also in order to rescale the power history during both the base irradiation in PWR for MTR irradiation.

VERDON Experimental Sequence

The VERDON thermal-hydraulic sequence is illustrated in Figure 5. Three main different phases can be established according to the temperature and atmosphere (Table 4).

Phase 1 and 2 are similar to VERCORS RT6, phase 3 is different (reducing for VERDON-1, oxidizing for RT6). Besides, one can note that there is a 1.5 hour troubleshooting resolution phase between phase 1 and phase 2. During this whole period, because of a technical problem, it was impossible to inject water, as it was initially defined, and consequently, to produce steam.

But, as the troubles occurred at relatively low temperature (less than 1000°C) the release of most of the FP had not begun and the impact of the temperature stabilization and slight decrease on FP release is negligible.

As explained above, some helium has been injected in the susceptor circuit (He Sus) during the whole sequence so that the over pressure between the experimental circuit and the susceptor circuit is maintained and the protection of the W susceptor tube regarding oxidation is guaranteed.

The criterion which is used to determine the end of the sequence is either to detect the fuel sample relocation or to reach the maximal temperature (T max = 2610°C). As illustrated in Figure 6, at the end of the sequence (1700°C<T<2610°C), no significant 140 La signal decrease was recorded by the gamma station. The end of the accidental sequence (shut down of HF power supply) was performed when the maximal temperature of 2610°C was reached. There is a total release of the iodine and cesium species measured at the end of the test, with an equivalent behavior, in term of release rate, between these two elements (Figure 7). The released fraction obtained at the end of the oxidation plateau is very high (around 60% of the initial inventory) with a start of release at ~1200°C. Beside, right before the beginning of release (1000°C<T<1200°C) a slight decrease of the "released fraction" of 137 Cs is detected (Figure 7).

FP release and fuel behavior

This is due to a cesium's axial and longitudinal migration inside the fuel pellet 5 .

The fission gases release kinetics (Figure 8) are characterized by a series of burst releases. In each case, the amplitude is greater for Xe than for Kr, in line with the expected Xe/Kr ratio. The first burst release starts at around 700-800°C (about 4% of Kr and 6 % of Xe). There is a second little burst release around 1000°C (~ 2% both). The following burst release starts at about 1100°C (16% of total Kr and 14% of total Xe). The main burst release occurs during the oxidation plateau at 1500°C (43% of total Kr and 47% of total Xe). A series of low burst releases occurs from 1800°C to the end of heating (20% of total Kr and 21% of total Xe). The last puff (15% of total Kr and 10% of total Xe) occurs at the beginning of the decrease of temperature. The fission gases release kinetics show a time lag between Kr and Xe at every burst end. This behavior can be a consequence of the zeolite contained in the May-Pack which yields to a separation of these two gases 6 . The Kr global release is slightly underestimated because of its limit of detection (about 1ppm), and its relative low concentration (compared to Xe). The loss of Kr signal, calculated between first and last burst release does not exceed 4% of total Kr measured. The loss of signal of the last puff, due to the end of measurement, can be evaluated up to 7% of total Xe and Kr measured (symmetrical burst end). The last puff probably indicates the complete drainage of the residual intragranular inventory of fission gases.

Figure 9 presents the Ba and Mo release kinetics up to 1500°C in order to allow a direct comparison with the VERCORS RT6 test (see section 5, discussion). A start of Mo release during the beginning of the so-called oxidation plateau at 1500°C with a very strong rate release kinetics up to approximately 40% of the initial inventory at the two third of the plateau is monitored. More or less symmetrically as compared to that of molybdenum, the release of barium does not start during the 1500°C plateau. Only a slight decrease of the "released fraction" of Ba is visible in Figure 9. This is probably due to barium axial and/or longitudinal migration inside the fuel pellet as it was observed in many VERCORS test (see section 2, [2]).

In order to illustrate the general behavior of the FP belonging to the four volatility categories, Figure 10 compares the release kinetics of Zr, Cs, I, Ba and Mo.

FP global behavior

Just before the loop dismantling and in order to get preliminary information on both FP release and FP final distribution inside the sample, the fuel sight gamma station is used as a longitudinal bench. The sample is then gamma scans inside the furnace. A comparison between the results obtained before and after the test is then performed according to a qualitative method  no or very low deposits respectively for the non (Zr, Nd) or low volatile (Nb, Np, Ru) FP. In the latter case the deposits are in the high temperature part of the sleeve (i.e. just above the crucible).

 Semi volatile FP present lot of deposits all along the sleeve with a significant retention in the sample.

 Volatile FP (I, Cs, Te) present no or a very little retention into the sample with low deposits in the extreme upper part of the sleeve just before the filter entrance.

As previously measured, the non volatile FP distributions do not evidence any significant fuel collapse. In fact, the general shape of the VERDON sample is found: two half un-irradiated pellets and the two PWR fuel pellets. The major modification lies on a global swelling of the sample of about 20% in length.

Finally, as observed during the VERCORS program for the volatile FP:

 limited cesium retention inside the PWR fuel is measured  Iodine is quasi only detected inside the un-irradiated half pellet.

Finally, as explained in above, all the loop elements are gamma scanned in order to perform the global FP balance. From a general point of view, one can note the very good consistency, in term of released fraction, obtained by the different methods (Figure 15): [START_REF] Leveque | The HEVA experimental programme[END_REF] differential between the measurement of the sample before and after the test (RF APS/AVS) and

(2) the sum of the deposited fractions. This point is probably one of the most important information deduced for this first VERDON test. In fact, this confirms the very good behavior of the loop during this type of accidental sequence. Moreover, the released fractions obtained by the on-line gamma fuel sight are also very consistent with the previous values (Table 5).

Discussion

The VERDON-1 results, in term of FPs and fuel behavior, presented in the previous part of this paper may be discussed according to three main axes: (1) fuel behavior with special focus on relocation temperature, (2) comparison with VERCORS RT6 test and finally (3) general FP volatility.

Fuel relocation

For the so-called "VERCORS HT and RT" series, the experimental facility had been totally refurbished with two complementary test loop configurations: the complex HT configuration [START_REF] Malgouyres | European working group "hot laboratories and remote handling" meeting[END_REF] (three tests), integrating the additional aim of studying FP transport in the primary system of a PWR and their potential interaction with the elements composing the neutron absorbers of a PWR (Ag, In, Cd and B, [START_REF] Pontillon | Fission product release and transport in severe-accident conditions: comparison between VERCORS HT1/HT2/HT3 experiments concerning UO 2 fuel in reducing and oxidizing conditions with and without control rod components[END_REF]), and the compact RT configuration (eight tests), more specifically focused on the release of low volatile FP and transuranium elements.

All these tests were conducted by bringing the fuel sample up to fuel relocation to quantify the entire volatility range of FP [2]. ; thus there is no great effect of the high burn up. Besides, whatever the test atmosphere conditions, the temperature at which the fuel loses its integrity is systematically inferior to both the melting point of un-irradiated UO 2 and the solidus temperature of the ZrO 2 -UO 2 eutectic [16].

Moreover, fuel collapse temperature seems to decrease in oxidizing conditions. This point is well highlighted by HT1, HT2 and HT3 tests which were performed on the same fuel section in reducing conditions for HT1 and HT3 and oxidizing conditions for HT2. The corresponding fuel collapse temperatures are approximately ~2300°C for HT1 and HT3 and ~2000°C for HT2.

This general behavior has been already discussed in details elsewhere [START_REF] Pontillon | Study of the active role played byUO2-ZrO2-PF interactions on irradiated fuel collapse temperature from VERCORS tests[END_REF], [START_REF] Barrachin | New modelling of the U-O-Zr phase diagram in the hyper-stoichiometric region and consequences for the fuel rod liquefaction in oxidising conditions[END_REF]. According to VERDON-1 test does not result in a global and/or strong relocation of the fuel sample at the end of the test (i.e. up to a temperature of 2610°C). This behavior may be connected, by analogy to the previous explanation, to the final atmosphere of the test (i.e. reducing conditions). However, the temperature "reaches" by the VERDON-1 sample is certainly one of the main points of the test, and at this stage, the exact origin of this behavior needs to be clearly demonstrated and more investigated [START_REF] Schwarz | PHEBUS FP: a severe accident research programme for current and advanced light water reactors[END_REF].

Comparison with RT6

As explained in introduction, the first part of the VERDON-1 test (i.e. up to the end of the oxidation plateau at 1500°C) was performed under the same atmosphere conditions compared to VERCORS RT6 test, and was conducted on a very similar UO 2 high burn up fuel, in order to check the continuity between VERCORS and the future VERDON data bases. In order to perform these analyses, Figure 9a and Figure 9b The released fraction obtained thanks to the data analysis presented in this paper are in good agreement with the VERCORS results and with what we can expect regarding FP behavior in VERDON-1 thermal-hydraulic condition (i.e. reducing/neutral at high temperature).

In more details, following information may be highlighted thanks to a general comparison with the VERCORS data base and in line with the above FP classification.

For all the VERCORS tests performed at temperature representative of a severe accident, the fission gas release was complete. The instantaneous fission gas (FG) release kinetics were characterized by successive burst releases at each temperature ramp. Besides, a higher release of long half-life FP ( 85 Kr) at low temperature (below 1200-1300°C) is monitored, due to the higher gas content inventory located at the grain boundaries for this type of gas [START_REF] Noirot | Fission Gas Inventory in PWR high burnup fuel: Experimental Characterisation and Modelling[END_REF],

compared to short half-life FP ( 133 Xe, 135 Xe) [START_REF] Pontillon | Fisson Gas Release From High Burn-up UO 2 Fuels Under Simulated Out-of-Pile LOCA Conditions[END_REF], as well as the final puff linked to the fuel melting and inducing the release of the ultimate gas fraction contained in the intragranular 7 It was shown that actinides can be subdivided into two categories. The first includes U and Np with released fraction that can reach 10%, with a behavior similar to the low-volatile category; the second (Pu) has very low released fraction, typically considerably less than 1%,with a behavior more-like non volatile FP bubbles. Between these two phenomena, fission gas release is driven by intragranular diffusion.

Results obtained for VERDON-1 are in perfect line with these observations. Beside, even though no general fuel collapse has been obtained, a very last puff is recorded at the end of the sequence where a complete release is measured.

Volatile FP have a high or even almost complete release for temperatures of around 2330°C. The nature of the test (fuel type, initial geometry, atmosphere at the end of the test, etc.) essentially affects the release kinetics of these species and has little effect on the released fraction once this temperature level has been attained during the test. I and Cs release rate are somewhat equivalent. There was a total release for iodine and almost total release for cesium in all the most severe VERCORS tests. In fact, it has been measured that a low but significant amount of cesium remained present in the sample for various tests, with released fractions of around 97-98%. Cesium retention in the corium was also identified as a result of the TMI2 accident, where associations with oxides stable at high temperatures were proposed: metal oxides of chromium (Cs 2 Cr 2 O 4 ) or iron (Cs 2 Fe 2 O 4 ), or silicates (Cs 2 Si 4 O 9 ) [START_REF] Hobbins | Fission product release from fuel under severe accident conditions[END_REF]. Apart the complete release of these FP, the strong fuel burn-up impact on release kinetics has been also highlighted. In order to illustrate this point, Figure 17 shows a comparison between the release kinetics of cesium during RT1 (considered as RT's reference test) and RT6 (high burn-up test).

The 137 Cs release kinetics during RT6 is much faster than for RT1, conducted in similar atmospheric conditions; for instance, at the end of the "oxidation plateau" (1500°C), the fraction of cesium released is approximately three times higher for RT6 and, throughout the test, the corresponding fractional release is at every moment greater. A similar increase of the release kinetics was observed for (1) MOX fuel, compared to UO 2 fuel, (2) "debris bed" configurations, compared to rod-like geometry, as well as for (3) oxidizing atmosphere instead of reducing.

Again the VERDON-1 results are very similar, with small retention of cesium inside the sample (<0.5%) and a faster release kinetics compared to a fuel with a moderate burn-up. For instance, the release was approximately of 60% at the beginning of the 1500°C plateau in VERDON-1 to be compared to around 20% at the same moment for VERCORS RT1 (4 cycles UO 2 fuel, Figure 17). Generally speaking, tellurium and antimony release are comparable and quasi total for all of the most severe VERCORS sequences. The main difference is in terms of quantities deposited in the hot zones of the experimental loop: much higher for antimony than tellurium. For VERDON-1, Sb and Te releases are respectively 76% and 96% with deposited fractions of ~50% and 32% (in the same order) along the sleeve. These results are in good agreement as far as the deposited fraction along the sleeve is concerned and also for the total release of Te. On the opposite, the released fraction of Sb is small. In fact, it has been noted that the Sb releases were lower for the RT grid tests (80-95%) compared to the VERCORS 3, 4 and 5 tests (97-100%). This behavior has been attributed to antimony retention in the corium formed after fuel relocation. Since VERDON-1 is characterized by a high temperature without fuel collapse, a stronger release should be expected. However, this point may be also due to an experimental artefact linked to the low detection, by gamma spectrometry, which characterizes these two FP. If it is the case, the released fractions obtained at this stage would correspond to the minimum of release. In other words, the "real" release would be higher. According to this line, if the detection limit calculated during the data treatment of the crucible zone is taken into account as a real FP detection for Sb (125 and 127 isotopes), an amount of around 5 to 10% in the crucible should be deduced from this approach, which would correspond to a global release (deduced from APS/AVS by the classical differential method) of 90-95%.

The behavior of the semi-volatile FP is characterized by releases that can be very high, in some cases as much as those of volatile FP, i.e. near total release, but with high sensitivity to the oxygen potential, and giving rise to significant deposits on the sleeve located above the fuel sample. The Mo release is enhanced in oxidizing conditions, on the opposite Ba release is enhanced in reducing conditions. Moreover, release kinetics seems to be faster for high burn up fuel. In fact, the results obtained for VERCORS RT6 show a significant increase in the fractional release compared to VERCORS 4 and 5 (Figure 18); for instance, at T=2000°C, 40% and 100% for Ba and Mo respectively for VERCORS RT6 instead of 0%(0%) and 10% (70-80%) for VERCORS 4 (and 5). VERDON-1 results confirm these observations with a very fast and high Mo release in the first part of the test (oxidizing H 2 O/H 2 atmosphere conditions). In the same time no Ba release is measured. On the other hand, during the second part of the test (reducing conditions), the release of Mo stops and the Ba release becomes very fast. After the sequence no drastic differences are measured with a slightly higher release of Ba compared to Mo (~65% and ~70% respectively for Mo and Ba) with deposits along the sleeve of ~19%(Mo) and 24%(Ba) of the initial inventory, which correspond to approximately 30-35% of the total release.

Low volatile FP have low, yet significant, released fraction of around 3% to 10% on average, but these values can attain 20-40% in the case of some FP under particular conditions, e.g. oxygen potential or high burn-up. In addition, the FP in this category are essentially deposited in the high temperature section of the test loop, i.e. close to the fuel. At this stage of the data analysis, the VERDON-1 results are also very consistent with these observations since no significant release has been measured by the fuel sight gamma station which is known to have a low sensibility which allows monitoring precisely release superior than 10% [START_REF] Ducros | Fission Product release under severe accidental conditions; general presentation of the program and synthesis of VERCORS 1 to 6 results[END_REF] of the initial inventory.

Zr and Nd do not have produced any detectable release during VERDON-1, as it has been the case for all the VERCORS tests.

Conclusion

This paper deals with the VERDON-1 test.

Fission products and nuclear fuel behavior under severe accident conditions -Part 1 :

 1 main lessons learnt from the first VERDON test Y. Pontillon*, E. Geiger, C. Le Gall, S. Bernard, A. Gallais-During, P.P. Malgouyres, E. Hanus and G. Ducros CEA, DEN, CAD, DEC, F-13108 Saint-Paul-lez-Durance, France.

  build a co-operative research program between teams involved in severe accident (SA) phenomenology all over the world (US-NRC, IRSN, CEA, EDF, PSI, European Commission, EACL, KAERI, …), based on separate-effect experiments and called "International Source Term Program (ISTP)". The results of these separate-effect experiments would allow improving models used for Source Term evaluation studies. Four main R&D research axes have been included in this program: (1) iodine study, (2) study of the boron carbide effect, (3) study of the air effect on fuel behavior and (4) study of the fission product release from the fuel.

4. 1

 1 FP releaseIn this part, the release kinetics, recorded by the fuel sight gamma station, is described according to the well-established FP classification (see discussion): non or low-volatile, volatile (including fission gases) and semi-volatile fission products.The release kinetics shown in Figure6aand Figure6billustrate that no significant release of Zr, La, Eu, Np and Ru is measured at this stage. Besides, no global relocation of the fuel sample is recorded by this gamma station (i.e. no global loss of the signal is recorded up to the end of the test at 2610°C).



  (i.e. the efficiency is considered equal to 1 in both cases). Comparison of the FPs distributions along the sample before (blue curve) and after (pink curve) the VERDON-1 test, obtained by gamma scanning of the sample inside the furnace, is shown in Figure11, Figure12and Figure13respectively for non or low-volatile ( 95 Zr,154 Eu, 103 Ru), semi-volatile ( 99 Mo and 140 Ba) and volatile FPs ( 137 Cs and 131 I).From a general point of view, the main information deduced from the on-line measurements (see section 4.1) is confirmed by the FP distribution along the sample: No significant release is measured, at this stage (and with this approach), for the so-called non or low volatile FPs, excepted for Ru where some little deposits are observed just above the crucible, No general degradation of the sample is highlighted by these post test gamma signals (i.e. for the non volatile FP), since the final shape of the sample (after the test) illustrated by the FP distribution, is more or less the same than that before the sequence, Strong and quasi total release respectively for semi-volatile and volatile FPs, and stronger release of Ba compared to Mo are measured.After these measurements, the sleeve is extracted from the furnace and gamma measured on the C4-gamma bench. The total gamma count as a function of the displacement along the sleeve is given in Figure14. The corresponding FP location is pointed out on this distribution. They are very consistent with the previous results and the well-known class of FP volatility, with:

Figure 16

 16 displays fuel collapse temperatures as a function of different selected VERCORS tests. Systematic fuel collapse has been detected for a temperature range of 2100-2350°C whatever the burn up from 47 GWd.t HM -1 to 70 GWd.t HM -1

  these data analysis in the UO 2+x composition domain, a new thermodynamic modelling of U-O phase diagram has been defined. An important consequence of this new optimization is that a liquid phase may appear in the O-UO 2 -ZrO 2 composition domain of the U-O-Zr phase diagram at 2330°C at atmospheric pressure (this temperature decreasing with increase of pressure, about 2230°C at 2 atm.). These temperatures can be associated with the temperature at which the fuel assembly could lose its integrity in oxidizing conditions and then with what was observed in some of the VERCORS tests (and quite differently from reducing test conditions) or in the PHEBUS tests.

Figure 1 .Figure 2 .Figure 3 .Figure 4 .Figure 5 .Figure 6 .Figure 7 .Figure 8 .Figure 9 . 1 :Figure 10 .Figure 11 .

 12345678911011 Figure 1. Schematic view of the VERDON laboratory

Figure 12 .

 12 Figure 12. Gamma scanning of the sample inside the VERDON furnace ( 99 Mo and 140 Ba) before (in blue) and after (in pink) the test.

Figure 13 .

 13 Figure 13. Gamma scanning of the sample inside the VERDON furnace ( 137 Cs and 131 I) before (in blue) and after (in pink) the test.

Figure 14 .Figure 15 .Figure 16 .Figure 17 .Figure 18 .

 1415161718 Figure 14. Total gamma count rate along the VERDON-1 sleeve

  For Ba and Mo the releases reach respectively approximately 0% and 40-45% in the two experiments. For volatile species, around 60% are obtained in the two considered experiments. These latter points are very important since they highlighted a perfect continuity between the two facilities in term of FP release[START_REF] Gallais-During | VERDON Laboratory : Performances of the Experimental LWR Severe Accident Device and First Results of Fission Products Release on High Burn-up UO2 Fuel[END_REF].

	5.3 FP volatility
	From a general point of view, analysis of the released fractions obtained during all the
	VERCORS program [2] made it possible to classify the FP into four categories of decreasing
	volatility: Volatile FP (including fission gases, iodine, cesium, antimony, tellurium, cadmium,

compare the release kinetics respectively for the semi volatile and volatile fission products during VERCORS RT6 and VERDON-1 tests. If one looks at the middle of the oxidation plateau, in both cases the general FP behavior (release rate and kinetics) are very similar.

rubidium and silver), FP such as molybdenum, rhodium, barium, palladium and technetium, Low volatile FP such as ruthenium, niobium, strontium, yttrium, lanthanum, cerium and europium and finally Non-volatile FP include zirconium, neodymium and praseodymium 7 .

  The main issue addressed by this first test concerns high burn up UO 2 fuel behavior -and corresponding fission product releases -under reducing conditions at very high temperature (up to 2600°C). Moreover, the first part of the test (i.e. up to the end of the oxidation plateau at 1500°C) has been performed under the same atmosphere conditions compared to VERCORS RT6 test, which was conducted with a very similar UO 2 high burn up fuel, in order to check the continuity between VERCORS and the future VERDON's data bases. This experiment has been performed in the new VERDON laboratory built at the Cadarache CEA center.During this VERDON-1 test, the good performances of the VERDON loop in terms of tightness, thermal-hydraulics, hafnia ceramics behavior, etc.… and of the gamma scanning and sighting have been clearly demonstrated. As a consequence it can be now asserted that the The results, in term of FP and fuel behavior, presented in the previous part have been discussed according to three main axes: (1) relocation temperature, (2) comparison with VERCORS RT6 test and finally (3) general FPs volatility.Generally speaking, the VERDON-1 test does not result in a global and/or strong relocation of the fuel sample at the end of the test (i.e. up to a temperature of 2610°C). This behavior must be connected to the final atmosphere of the test (i.e. reducing conditions).According to the released fractions measured by on-line gamma station and thanks to the information obtained via pre and post test gamma scanning, the FP general classification, in

	The comparison with VERCORS RT6 has been possible and conclusive. Similar FP release
	kinetics at 1500°C, high burn up effect on release kinetics and atmosphere effect on Mo, Ba
	release are measured. The VERDON loop is thus qualified in "release configuration", and
	VERDON is in continuity with VERCORS experiments.

relation to their released fractions and specific behavior, is again obtained with: (1) volatile FP (fission gases, iodine, cesium, tellurium, antimony) with an almost total release; (2) semivolatile FP (molybdenum and barium), with high sensitivity to oxidizing-reducing conditions and significant released fractions; (3) FP that are low or no volatile (ruthenium, europium, niobium, cerium, zirconium, neodymium).

Table 1 . VERCORS 1 to 6 test matrix parameters

 1

	Test	VERCORS 1	VERCORS 2	VERCORS 3	VERCORS 4 VERCORS 5 VERCORS 6
	Date of test	11-1989	06-1990	04-1992	06-1993	11-1993	06-1994
	Fuel
	PWR irradiation	Fessenheim	Bugey	Bugey	Bugey	Bugey	Gravelines
	Fuel burn-up (GWd/tU)	42,9	38,3	38,3	38,3	38,3	60
	Re-irradiation	Siloe	Siloe	Siloe	Siloe	Siloe	Siloe
	Test conditions
	Max fuel temperature (K)	2130	2150	2570	2570	2570	2620
	Atmosphere (end of test)	Mixed H 2 O+H 2 Mixed H 2 O+H 2	Mixed H 2 O+H 2	Hydrogen	Steam	Mixed H 2 O+H 2
	Last plateau duration (min)	17	13	15	30	30	30
	Steam flow rate (g/min)	0,15	1,5	1,5	1,5 -0	1,5	1,5
	Hydrogen flowrate (g/min)	0,003	0,027	0,03	0,012	0	0,03

Table 2 . VERCORS HT-RT test matrix parameters

 2

	VERCORS tests	HT 1	HT 3	HT 2	RT 1	RT 2	RT 5	RT 4	RT 3	RT 7	RT 6	RT 8
	Date of test	June 1996	June 2001	April 2002	March 1998 April 1998	December 1998	June 1999	November 1999	April 2000	September 2002	November 2002
	Fuel	UO 2	UO 2	UO 2	UO 2	MOX	UO 2	UO 2 /ZrO 2 debris bed	UO 2 debris bed	MOX	UO 2	UO 2
	Burnup (GWd/tU)	47	~47	~47	47	41	60	3 cycles	3 cycles	3 cycles	6 cycles	6 cycles
	Re-irradiation	SILOE	OSIRIS	OSIRIS	No	No	OSIRIS	No	OSIRIS	OSIRIS	OSIRIS	OSIRIS
	Max fuel temperature (K) / Fuel collapse	2900 / 2500 2750 / 2500 2600 / 2300	2570	2440	Fuel collapse	Fuel collapse	Fuel melting	Fuel melting	Fuel melting	Fuel melting
	H 2 (mg/s)	0,2	0,2	0	0,45	0,45	0,45	0,4	1,25	0,2	0,45	0
	H 2 O (mg/s)	0	0	25	25	25	25	14,6	1,25	0	25	0
	Air (mg/s)	0	0	0	0	0	0	0	0	0	0	0,8
	Main objective	H 2 atm., high temperature, HT reference test	Boric acid and SIC injection	Boric acid and SIC injection	RT reference test	MOX fuel	High Burnup	Phebus FPT4 support	Fuel volatilization	MOX fuel	High burn up fuel	High burn up fuel / air injection

Table 3 : FP's initial inventory of the VERDON 1 sample, comparison between gamma spectrometry measurements and Cesar 5.1 calculations, with

 3 (M/C)1 and (M/C)2

respectively for OSIRIS and PWR power history.

				UO2 PWR
	FP	Half-life	Measurements	Calculations	Calculation*(M/C) (M/C)1, (M/C)2	M/C
	Zr95	63.98d	3,54E+16	3,39E+16	3,60E+16	0,98
	Nb95	65,00d	1,77E+15	1,58E+15	1,68E+15	1,06
	Mo99	2,75d	1,81E+16	1,67E+16	1,77E+16	1,02
	Ru103	39,3d	3,96E+16	3,89E+16	4,13E+16	0,96
	Ru106	1,017y	3,83E+16	3,65E+16	3,80E+16	1,01
	Sb125	2,76y	3,95E+16	1,03E+17	1,07E+17	0,37
	Sb127	3,85d	1,19E+15	1,35E+15	1,43E+15	0,83
	Te132	78,20h	1,58E+16	1,47E+16	1,56E+16	1,01
	I131	8,02d	1,76E+16	1,64E+16	1,74E+16	1,01
	I133	20,80h	7,13E+15	6,57E+15	6,97E+15	1,02
	Cs134	2,07y	5,49E+17	5,32E+17	5,54E+17	0,99
	Cs137	30,17y	1,16E+20	1,11E+20	1,16E+20	1,00
	Ba140	12,8d	3,30E+16	3,08E+16	3,27E+16	1,01
	La140	1,68d	4,54E+15	4,03E+15	4,28E+15	1,06
	Ce141	32,50d	3,74E+16	3,36E+16	3,57E+16	1,05
	Ce143	33,00h	8,00E+15	7,32E+15	7,77E+15	1,03
	Nd147	11,00d	1,08E+16	1,16E+16	1,23E+16	0,87
	Eu154	8,80y	1,14E+18	2,37E+18	2,47E+18	0,46
	Eu156	15,19d	3,13E+15	2,86E+15	3,04E+15	1,03
	Np238	2,12d	6,61E+15	5,91E+15	6,27E+15	1,05

Table 5 : Comparison between released fractions obtained by the differential approach (before versus after test) and by on-line gamma fuel sight

 5

	FP	RF (APS/AVS) crucible	RF fuel sight Kinetics
	Kr85	100,0%	ND
	Zr95	0,3%	0,0%
	Nb95	0,0%	<5-10%
	Zr97	0,9%	0,0%
	Mo99	63,7%	60,0%
	Ru103	1,3%	0,0%
	Sb125		ND
	Sb127		61% à 2120°C
	Te132	95,0%	67% à 1670°C
	I131	93,7%	100,0%
	I133	ND	100,0%
	XE133	100,0%	ND
	Cs134	99,9%	100,0%
	Cs137	99,4%	100,0%
	Ba140	76,4%	~75%
	La140	0,0%	0,0%
	Ce143	0,2%	0,0%
	Nd147	0,0%	0,0%
	Eu154	-1,1%	0,0%
	Np238	2,8%	0,0%
	Np239	2,1%	0,0%

The maximal temperature that these joints can sustain is about 300°C and for higher temperature, the tightness of the furnace would not be guaranteed. For this reason, these joints need to be adequately cooled by a dedicated cooling circuit located inside the bases, facing the joints

At least 10% release have to be recorded by this station to guarantee a significate value, particularly as the changes in the object geometry measured during heating (swelling, fracturing, then fuel collapse, etc.) significantly complicate the quantitative use of the measurement, just like the axial migration of the FP.

This initial inventory prior to the sequence is often determined in two stages: before the re-irradiation in the experimental reactor to precisely measure the FP with long half-lives, then just after this re-irradiation to measure the FP with short half-lives.

A similar approach is performed for the un-irradiated half pellet, the corresponding measured amount is then included in the total FP initial inventory.

This FP axial and/or longitudinal migration inside the fuel pellet is seen as an over-concentration by the gamma station, resulting in an apparent decrease of the "released fraction" as it has been previously observed during the VERCORS programme

This zeolite "pre-column" effect was previously seen on calibration gas Kr and Xe during commissioning tests before VERDON-1 but was not quantified


E Geiger

Le Gall

A Gallais-During

Yves Pontillon

J Lamontagne

Eric Hanus

Fission products and nuclear fuel

come

Within the frame of the International Source Term Program, VERDON program aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a Light Water Reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO 2 fuel and FP release at very high temperature (≈ 2600°C) under reducing atmosphere. Post-test qualitative and quantitative characterizations of the VERDON-1 sample led to the proposal of a scenario explaining what might have occurred during the experimental sequence. Hence the fuel and the cladding might have interacted leading to the melt of UO 2 -ZrO 2 system. Besides no relocation was observed during the test, it might have been imminent.

INTRODUCTION

Within the frame of the International Source Term Program, different tests were performed in the VERDON laboratory at the LECA-STAR facility (CEA Cadarache). Their objective was to quantify the source term of LWR in case of a hypothetical severe accident. More precisely, the VERDON-1 test was devoted to high burn-up UO 2 fuel behavior and FP release under reducing conditions at very high temperature (up to 2610°C).

In this work, a detailed characterization including Scanning Electron Microscopy (SEM) and Optical Microscopic (OM) observations as well as Electron Probe Micro-Analyses (EPMA) of the VERDON-1 sample before and after the test is given. A special focus is made on the different chemical phases found into the fuel and a scenario explaining the evolution of the fuel during the test is proposed. The VERDON-1 experimental circuit and the progress of the accidental thermal-hydraulic sequence are briefly summarized as it has been extensively discussed in a first article [START_REF] Pontillon | Fission products and nuclear fuel behavior under severe accident conditions -Part 1: Main lessons learnt from the first VERDON test[END_REF].

FUEL SAMPLES CHARACTERISTICS

VERDON-1 father rod

The father rod from which the VERDON-1 sample was extracted has been irradiated for six cycles in a Nuclear Power Plant (NPP) operated by EDF. The mean burn-up of the rod was calculated by EDF to be 67.535 GWd.t U -1 . Its characteristics are summarized in Table 1. SEM and EPMA were performed at the periphery of the fuel in contact with the cladding, at half diameter from the periphery and at the center of the pellet. This allowed an accurate description of the overall fuel sample before the VERDON-1 test.

SEM observations were performed using a Philips XL30 device. Large field acquisitions were performed using the ADDA SIS system and the AnalySIS software was used for images analyses.

Concerning EPMA, a shielded CAMECA SX100R device was used for acquisition and exploitation of the measurements. Incident electron beam characteristics were set at 20 kV and 200 nA.

At the periphery, the High Burn-up Structure (HBS) has been observed [START_REF] Noirot | Detailed characterisations of high burn-up structures in oxide fuels[END_REF]- [START_REF] Spino | Detailed characterisation of the rim microstructure in PWR fuels in the burn-up range 40-67 GWd/tM[END_REF]. It extends in a ring up to 60 µm from the periphery. This restructuration is characterized by the presence of small round and polyhedral sub-grains and a dense concentration of pores which sizes vary from 0.4 to 2 µm. A fractography of this region is presented on Figure 1. EPMA were performed on a polished cross-section of a pellet, from the periphery up to the center. Close to the periphery, higher Pu and Nd contents, 4.3 and 1.5 wt% respectively, have been noticed coinciding with the HBS. Their concentration decreases down to 1 and 0.7 wt% respectively around 800 µm from the periphery up to the center of the sample (Figure 2). According to Nd concentration, the local burn-up was estimated to be 150 GWd.t U -1 in the HBS region. Small precipitates of diameter from 0.2 to 1 µm were found mainly between the center and 2500 µm from the edge of the pellet (Figure 3). Their density is more elevated at the periphery and in the center than in the half diameter region of the sample. As shown thanks to quantitative profiles obtained by EPMA (Figure 4), these precipitates are composed of Mo being the main constituent, Ru and Pd. They are common in irradiated nuclear fuels [START_REF] Kleykamp | The chemical state of the fission products in oxide fuels[END_REF].

Beside the elevated density of metallic precipitates in the center of the fuel pellets, an important inter-granular porosity was also observed with a pores size of about 1 µm. The fractography and Cs X-ray map presented on Figure 5 reveal the presence of cesium precipitates at half-diameter of the pellet. No association was observed between cesium and molybdenum or tellurium. No such precipitates were observed in the same region during the EPMA analysis of a polished sample. It was assumed that these precipitates had been eliminated during the preparation of the sample. Small bubbles of xenon, not shown in this paper, were also observed in this region, with a diameter between 0.1 and 0.2 µm.

Xe and Cs quantitative radial profiles are presented on Figure 6. The Xe creation quantity has been estimated from Nd quantitative analyses fitted with the following equation [START_REF] Blay | FX0GAC/E04/4034 -Examen au MEB et à la microsonde d'un combustible REP UO2 ALIX irradié 6 cycles dans GRAVELINES 5[END_REF]. A correction corresponding to the creation ratio Xe/Nd has been applied.

/ /

The Xe signal is lost both at the periphery, coinciding to the HBS zone, and in the center of the sample. These gaps can be explained on the one hand by the release of gaseous Xe during irradiation in the reactor and on the other hand by the escape of gaseous Xe during polishing. On the contrary, Cs content is higher at the periphery (0.75 wt%) and decreases down to 0.4 wt% in the center. As most FP, their concentration is higher in the periphery of the fuel (around 0.64 and 1.15 wt% respectively) and decreases to around 0.2 and 0.58 wt% respectively 400 µm away from the periphery. No particular association between Ba and Zr was observed contrary to what was expected, BaZrO 3 being considered to be the most stable chemical state [START_REF] Kleykamp | The chemical state of the fission products in oxide fuels[END_REF].

The VERDON-1 sample

The VERDON-1 sample was composed of two irradiated fuel pellets in their original M5 cladding selected from the former UO 2 father rod. Two half pellets of un-irradiated depleted UO 2 were placed at each end of the sample and held in place by crimping the cladding. The cladding was not fully sealed, allowing an interaction with the surrounding atmosphere. The VERDON-1 sample configuration before the test is presented on Figure 8. The local burn-up of the fuel section was calculated to be 72 GWd.t U -1 .

Before the test, the sample has been re-irradiated at low linear power (10 W.cm -1 ) in OSIRIS material testing reactor (CEA Saclay) for about ten days. The objective was to recreate the short half-life FP such as 99 Mo, 132 Te, 133 I, 131 I, 140 Ba, etc.

VERDON-1 thermal-hydraulic sequence

The experimental loop has been extensively described in [START_REF] Pontillon | Fission products and nuclear fuel behavior under severe accident conditions -Part 1: Main lessons learnt from the first VERDON test[END_REF], [START_REF] Ducros | Use of Gamma spectrometry for measuring fission product releases during a simulated PWR severe accident: Application to the VERDON experimental program[END_REF]- [START_REF] Pontillon | Behaviour of fission products under severe PWR accident conditions VERCORS experimental programme-Part 1: General description of the programme[END_REF] so only the main characteristics are recalled in this section.

The release configuration of the VERDON-1 loop consists mainly of an induction furnace where the sample is heated, a fluid injection system (He, H 2 , steam, air …), an aerosol filter to trap FP under aerosol forms, a May-Pack filter to trap gaseous iodine species, a condenser to recover steam and a final safety filter. Three on-line γ-spectrometry stations enable release and deposit kinetics measurements during the thermal-hydraulic sequence.

The VERDON-1 test was carried out 64 hours after the re-irradiation of the fuel. Its thermal-hydraulic sequence consisted in three main phases which differ by the composition of the atmosphere (Figure 9):

• The first phase was a 400°C plateau maintained during one hour under neutral atmosphere; • The second phase began around 750°C under mixed steam-H 2 oxidizing atmosphere. The steam and H 2 injection flow rates (25 mg/s and 0.45 mg/s respectively) were maintained during nearly one hour at 1500°C in order to reach complete cladding oxidation; • A reducing atmosphere was imposed in phase 3, from 1500°C to the end of the test with a molar ratio H 2 /H 2 O of 10. The very end of phase 3 (from 2400°C to 2500°C) was performed with a decreasing H2 flow rate and finally, water injection was stopped during the last part of the test (up to 2600°C). The atmosphere was thus variable from reducing to neutral at the end of the test. The end of the sequence was defined by the first of these two criteria: the detection of the fuel sample relocation or a maximal temperature of 2600°C. No apparent relocation has been detected so the end of the sequence (shut down of HF power supply) was performed when the maximal temperature of 2600°C was reached. After the test, the fuel sample was recovered from the furnace. Though the fuel degradation was quite advanced indicating an imminent relocation of the fuel, both irradiated pellets had almost kept their original geometry (Figure 10). The sample was brittle as a consequence of the test and a partial melting of the fuel pellets and some missing pieces were observed. As indicated in Table 1, the two original pellets measured together 27.8 mm length, with a mean diameter of 8.086 mm. After the test, the length of the fuel element was approximately 31 mm with a diameter of 8.6 mm. This swelling is due to the increase of internal pressure and mechanical strains caused by the thermal-hydraulic sequence. This is probably the origin of the cracks observed into the fuel.

Characterizations setup

After the test, the two irradiated fuel pellets were separated. Each pellet was prepared in a different way: sample A was embedded in a Wood metallic alloy (Bi/Sn) and sample B in Epoxy resin. The samples in their respective embedding were polished up to the half of their diameter.

OM observations were performed using a LEICA DM RXA2 device adapted for hot cell work. Image acquisition was done with a LEICA DFC 320 camera on the SIS acquisition system.

SEM observations were performed in secondary electron mode using a Philips XL30 device. Large field acquisitions were performed using the ADDA SIS system and the AnalySIS software was used for images analyses.

Concerning EPMA, a shielded CAMECA SX100R device was used for acquisition and exploitation of the measurements. Incident electron beam characteristics were set at 20 kV and 200 nA.

In order to perform SEM analyses on sample B, a carbon coating was required to improve its electric conductivity. Though given their brittleness half a pellet was destroyed during the cutting process. Samples A and B were submitted to post-test examinations in the LECA-STAR facility (CEA Cadarache). SEM and EPMA were performed on sample A and a

Microstructure analyses

On these macrographs, an advanced state of degradation of the samples can be observed. This means that the relocation of the fuel was imminent at the end of the VERDON-1 test. Three main areas are observed on the Figure 10: • A predominant dark-grey zone with elevated porosity,

• A light-grey one with moderated porosity,

• A second light-grey phase with reduced porosity.

Representative images of these 3 regions are presented on Figure 11. The elevated-porosity region is characterized by an elevated number of small round pores and big irregular ones. These last may have been produced by the coalescence of the smallest ones. Furthermore, many white precipitates are present in the regions of elevated and moderated porosity. Regarding to the region of reduced porosity, a few big size precipitates were found (see the X-ray map on Figure 14).

A detailed image analysis gave the size distribution of pores and white precipitates in the different regions and their contribution to the surface. The surface contribution of the pores in the elevated porosity region almost doubles compared to the moderated porosity one: 42% and 25% respectively. Almost all the porosity is due to pores of size superior to 4 µm, in both regions. The total number of pores smaller than 4 µm is four times more important in the elevated porosity region than in the moderated porosity one.

Regarding the white precipitates, their contribution to the total surface is almost the same in the moderated porosity region than in the elevated porosity one: around 0.5 and 0.4%, respectively. The precipitates observed in the elevated porosity region are slightly smaller, with a mean Equivalent Circle Diameter (ECD) of 0.8 µm, than those from the moderated porosity region, with an ECD of 1.1 µm. Secondary electrons SEM (SE-SEM) images representative of the elevated and low porosity regions in samples A and B are presented on Figure 12. Fuel grains were clearly identified inside the pores of both regions, where also small round precipitates were observed. These precipitates, corresponding to the white precipitates observed by optical microscopy, have a size ranging between 1.9 and 3.2 µm. They are similar to those already observed in the father rod identified as metallic precipitates of Mo, Pd and Ru, but bigger.

B-1 OM B-2 OM B-3 OM

Chemical analyses

EPMA were performed on Sample A, as indicated on Figure 10 (radial dashed line). In this paper, the results of the transversal analyses are presented, since no particular radial or longitudinal concentration tendencies have been observed.

As observed on Figure 13, an important quantity of Zr was found in the reduced porosity regions. This result highlights the chemical difference between the elevated, moderated and low porosity regions: the lower the porosity, the higher the zirconium content.

A slight increase in Pu and Nd concentrations in the reduced porosity regions could be noticed, corresponding to the concentration profile of Zr. In the high porosity region, Pu and Nd concentrations were not modified after the test (Figure 13). On the contrary, Pu and Nd concentrations in the periphery decreased compared to that observed in the father rod quantitative analyses (Figure 2). Ru, Mo and Tc are observed in the same precipitates. However, it is important to notice that Bi which is one of the metallic embedding components used in the preparation of this sample can interfere with Tc signal.

Metallic precipitates

In accordance with the previous EPMA results, the region which presented the lowest quantity of pores was highly concentrated in Zr while the U content seems to be reduced compared to the most porous regions. To quantify this phenomenon, EPMA has been performed along the line scan 1 (Figure 14) crossing the three regions. This line started in the most porous region and ended in the center of the region highly concentrated in Zr without pores. The main results of this analysis are displayed on According to the profiles observed on Figure 15, U concentration was around 85 wt%, in the elevated porosity zone and gradually decreases down to 75 wt% in the moderated porosity region. Once in the reduced porosity region, it drops to 55 wt%. Regarding O, its concentration increases from 11.8 wt% in the high porosity region up to about 14 wt% in the center of the reduced porosity one. The same behavior is observed for Zr and Pu which concentrations increases from about 1 and 1.2 wt% to about 23 and 1.7 wt% in the center of the crack, respectively. These results were also confirmed by quantitative analyses performed along the A2-EPMA and A2'-EPMA lines (Figure 10). In two regions of moderated porosity located at the periphery of the fuel pellet originally in contact with the cladding. Zr concentration is elevated and a high content of Pu is also present.

DISCUSSION ON U-ZR INTERACTION

The post-test examination of the VERDON-1 sample has highlighted a change of microstructure after the test. Three zones have been observed with different porosity. Lowporosity regions are rich in Zr and Pu and depleted in U which is characteristic of a U-Zr interaction.

To go further in the understanding of this phenomenon, a sample composed of an irradiated fuel pellet in its original M5 cladding was selected from the same UO 2 father rod as the VERDON-1 sample. It was submitted to an annealing treatment consisting in two temperature plateaus at 300 and 1200°C during 15 minutes and a third one at 1500°C during 25 minutes under reducing atmosphere (Ar + 4% H 2 ). The test was carried out in the MERARG experiment loop [START_REF] Pontillon | Fission gas release under normal and off-normal conditions: New analytical device implemented at the CEA Cadarache[END_REF]. After the annealing test, the sample ("intermediate state sample" in the following) was embedded in an epoxy resin and polished down to half his diameter.

The evolution of the porosity into the fuel clearly indicates that the pores interconnection process began around 1500°C and was enhanced as the temperature increased (Figure 16). This phenomenon could explain the main burst release observed for Xe and Kr during the oxidation plateau. Indeed, as the pores interconnected, paths were created in the whole sample for gaseous FP to diffuse more rapidly in the sample and release from the fuel [START_REF] Gall | Fission products and nuclear fuel behavior under severe accident conditions -Part 3: Fission products speciation in the VERDON-1 sample[END_REF].

An evolution of the ZrO 2 layer can also be observed as it comes closer and interact with the fuel when the temperature increases. Quantitative results and X-ray maps indicated that there was indeed an interaction between Zr from the oxidized cladding and UO 2 from the fuel: a (U-Zr)O 2 mix melted and penetrated through the cracks and the pores, carrying FP present in the periphery of the fuel (HBS structure). This would explain the elevated concentrations of Pu and Nd observed in the moderated and reduced porosity regions. The behavior of these two non-volatile FP agrees with the results of VERCORS test [START_REF] Pontillon | Behaviour of fission products under severe PWR accident conditions VERCORS experimental programme-Part 1: General description of the programme[END_REF], [START_REF] Pontillon | Behaviour of fission products under severe PWR accident conditions: The VERCORS experimental programme-Part 2: Release and transport of fission gases and volatile fission products[END_REF], [START_REF] Pontillon | Behaviour of fission products under severe PWR accident conditions. The VERCORS experimental programme-Part 3: Release of low-volatile fission products and actinides[END_REF].

In order to determine whether ZrO 2 melted under VERDON-1 test conditions or a ZrO 2 -UO 2 or Zr-UO 2 interaction took place, thermodynamic estimations were made. As indicated in Part 3.1, during the third stage of the test, the fuel sample was submitted to a reducing atmosphere consisting of a H 2 /H 2 O mixture with a molar ratio 10:1, up to 2400°C. Assuming that the fuel was in equilibrium with the atmosphere and thus its oxygen potential µO 2 was imposed by it, the Wheeler and Jones equation can be used [START_REF] Wheeler | Thermodynamic and composition changes in UO2±x (x< 0.005) at 1950 K[END_REF] Considering the experimental values, µO 2 =-291 kJ.mol -1 .K -1 was obtained. According to the Ellingham diagram [START_REF] Kleykamp | The chemical state of the fission products in oxide fuels[END_REF] presented on Figure 17, the reduction of ZrO 2 to Zr would take place at less than -600 kJ.mol -1 for a temperature of 2400°C. This means that the reducing character of the VERDON-1 test atmosphere was not strong enough to reduce the oxidized cladding.

Since the maximal temperature at the end of the test was lower that the pure ZrO 2 melting temperature (around 2720°C), an interaction ZrO 2 -UO 2 may have taken place. A similar behavior has already been observed in the VERCORS program, where the fuel collapsed at temperatures 230°C lower than the expected UO 2 melting temperature [START_REF] Pontillon | Behaviour of fission products under severe PWR accident conditions VERCORS experimental programme-Part 1: General description of the programme[END_REF]. According to the pseudo binary ZrO 2 -UO 2 phase diagram presented on Figure 18, extracted from [START_REF] Piluso | The UO2-ZrO2 system at high temperature (T > 2000 K): importance of the meta-stable phases under severe accident conditions[END_REF], an indifferent point (the analogue to liquid-vapor azeotrope) exists at a molar fraction of x ZrO2 =0.58. Quantitative analyses performed after the test helped determining the XRD experimental data, not shown in this paper, seemed to agree with this composition.

Thus, it is assumed that a reaction between UO 2 and ZrO 2 from the cladding took place. When a composition matching (or close to) the indifferent point was attained, the solution melted and penetrated into the pellets through the cracks, carrying FP normally located in the HBS, such as Pu and Nd.

CONCLUSION

The VERDON-1 test was devoted to high burn-up UO 2 fuel behavior and FP release under reducing conditions at very high temperature (up to 2610°C). The sample recovered after the VERDON-1 test was characterized using OM, SEM and EPMA techniques. An intermediate test has been performed under reducing atmosphere up to 1500°C in the MERARG experimental loop. An irradiated fuel sample, obtained from the same father rod as the VERDON-1 sample was used in this test and has been characterized. This was the first time that these types of analyses were performed at the CEA Cadarache on a sample submitted to severe accidents conditions.

Characterizing the chemical phases in the fuel and their evolution allows understanding both the fuel and FP behavior during a severe accident. These results can be used for the improvement of mechanistic codes (such as the MFPR code developed by the IRSN) which would enable a better assessment of the source term for a severe accident.

The comparison between the intermediate state sample and VERDON-1 sample tends to prove that pores interconnection took place around 1500°C and end to the formation of three different regions: (1) a first region presenting an elevated density of irregular interconnected pores and small precipitates, (2) a second region presenting a moderated density of pores and also small precipitates, and (3) a third region presenting a low porosity with big size pores and a few isolated big size precipitates. These three regions differed greatly in the Zr content, increasing as the porosity decreases. At the same time, Pu and Nd presented similar concentrations profile than those of Zr.

Indifferent point

From these results, the following evolution is assumed to have taken place during the VERDON-1 test. First, the swelling of the samples due to an increase of internal pressure and mechanical stresses caused by the thermal-hydraulic sequence produced many cracks into the fuel. Being in contact with the pellet, ZrO 2 from the oxidized cladding diffused into the fuel attaining concentrations close to the eutectic point of the pseudo binary UO 2 -ZrO 2 system. It melted at the end of the test and penetrated into the fuel through the cracks. Many heavy elements that were located in the periphery of the pellet (such as Pu and Nd) were also dissolved and transported into the melt. ZrO 2 and other elements kept diffusing into the fuel matrix, leading to the three microstructures observed at the end of the test.

Further characterizations of the VERDON-1 sample are presented in the third part of this article [START_REF] Gall | Fission products and nuclear fuel behavior under severe accident conditions -Part 3: Fission products speciation in the VERDON-1 sample[END_REF]. These analyses are focused on the evolution of FP inside the fuel during the VERDON-1 test.

Fission products and nuclear fuel behavior under severe accident conditions Part 2 :

 2 Fuel behavior in the VERDON-1 sample E. Geiger*, C. Le Gall, A. Gallais-During, Y. Pontillon, J. Lamontagne, E. Hanus, G. Ducros CEA, DEN, DEC, F-13108 Saint-Paul-lez-Durance, France

Figure 1 :

 1 Figure 1: Fractography of the pellet periphery, VERDON-1 father rod

Figure 2 :

 2 Figure 2: Pu and Nd radial EPMA quantitative profiles, VERDON-1 father rod

Figure 3 :Figure 4 :

 34 Figure 3: SEM image of Mo, Ru, Pd, Tc and Rh metallic precipitates, VERDON-1 father rod 0 0,5 1 1,5 2 2,5 3 3,5 4 4,5

Figure 5 :Figure 6 :

 56 Figure 5: a) SEM fractography and b) Cs X-ray map, half-diameter, father rod

Figure 7 :

 7 Figure 7: Ba and Zr radial EPMA quantitative profiles, VERDON-1 father rod

Figure 8 :

 8 Figure 8: The VERDON-1 sample

Figure 9 :

 9 Figure 9: VERDON-1 thermal-hydraulic sequence

  metallographic analysis (optical microscope) was made on sample B, as well as SEM analyses. Only the most representative images obtained by each technique are presented below. Their location in the samples is presented on the macrographs of samples A and B (Figure10).

Figure 10 :

 10 Figure 10: Micrograph of the VERDON-1 fuel pellets after the test, Sample A (left) and Sample B (right) after preparation

Figure 11 :

 11 Figure 11: Optical Microscopy of region of elevated (B-1 OM), moderated (B-2 OM) and low porosity (B-3 OM), sample B

Figure 12 :

 12 Figure 12: SEM-SE, Region of elevated (A-1 SEM-SE) and reduced porosity (B-3 SEM-SE), Samples A and B

Figure 13 : 1 Figure 14 :

 13114 Figure 13 : Zr, Ba, Pu and Nd transversal EPMA quantitative profiles, sample A-EPMA Zone A' on Figure 10 was chosen to perform X-ray maps because the three regions of elevated, moderated and reduced porosity were present. Images of Zone A' presented on Figure 14 correspond to a SE image and U, Zr, O, Ba, Ru, Mo and Tc X-ray maps.

Figure 15 :

 15 Figure 15: U, Zr, O, Pu and Nd EPMA quantitative analyses. Line scan 1 in Zone A' Sample A

Figure 16 :

 16 Figure 16 : SEM images of the periphery of the (a) father rod sample (BSE), (b) the intermediate state sample (BSE) and (c) the VERDON-1 sample (SE)

  , Zr and O at the center of one crack. The molar fractions measured were 0.168; 0.183 and 0.636, for U, Zr and O respectively. They have been calculated from the measured mass fractions considering a phase containing Zr, U, O, Pu and Nd. With these results a Zr/(U+Zr) ratio of 0.52 which is not far from the indifferent point composition, was obtained. Its melting temperature (2580°C) and the temperature reached by the end of the VERDON-1 test (2600°C) were close enough to form a liquid solution.

Figure 17 :Figure 18 :

 1718 Figure 17: Ellingham Diagram [5]

Table 1 :

 1 VERDON-1 father rod characteristics

	Cladding material	Outer diameter of clad (mm)	Inner diameter of clad (mm)	Fuel	Pellets mean height (mm)	Pellets mean diameter (mm)
	M5	9.50	8.25	UO 2	13.65	8.086

  :

	μ	.	-479070 4.184 × " 8.86$ %&"' -4.42 9.152 % * +, -. +, -	/

AKNOWLEDGEMENT

This work was performed in the frame work of a collaboration including the CEA, EdF and IRSN. The authors are indebted to T. Blay, L. Brunaud, P. Delion, L. Fayette, K. Hanifi, J. Noirot, S. Reboul and I. Zacharie-Aubrun from the LEMCI's teams working in the CEA LECA-STAR hot-cells facility at Cadarache, for their huge contribution to this work. We also greatly thank S. Clément, M. Pontillon and G. Volle for their work on the MERARG loop.