M Dumerval

Q Houmaire

J C Brachet

H Palancher

J Bischoff

E Pouillier

BEHAVIOR OF CHROMIUM COATED M5 CLADDINGS UPON THERMAL RAMP TESTS UNDER INTERNAL PRESSURE (LOSS-OF-COOLANT ACCIDENT CONDITIONS) (FOR POSTER SESSION)

Behavior of chromium coated M5 claddings upon

thermal ramp tests under internal pressure (loss-of-coolant accident conditions)

INTRODUCTION AND EXPERIMENTAL PROCEDURE ("EDGAR" TESTS)

Chromium coated zirconium based fuel claddings are developed and studied within the CEA-Framatome-EDF joint research program as an "Enhanced Accident Tolerant Fuel" (EATF) cladding short-term concept for nuclear light water reactors. In addition to their already demonstrated improved resistance to High Temperature (HT) steam oxidation [START_REF] Brachet | Behavior of Cr-coated M5™ claddings during and after high temperature steam oxidation from 800°C up to 1500°C[END_REF], it has been previously shown, for both Zircaloy-4 and M5 substrates, that Cr-coatings, even with a thickness limited to 10-15µm, can have a positive impact on the HT isothermal creep and burst behavior of the cladding under internal pressure [START_REF] Brachet | Behavior under LOCA conditions of Enhanced Accident Tolerant Cr-Coated Zircaloy-4 Claddings[END_REF]- [START_REF] Brachet | Behavior of Cr-coated M5™ claddings under LOCA conditions[END_REF]. The results obtained so far showed that in some conditions the Cr coating had a HT strengthening effect associated with significantly smaller clad ballooning and/or rupture opening. Such a behavior, if confirmed for more realistic LOCA dynamic thermal ramp conditions, should be of interest to mitigate nuclear fuel sub-assembly flow blockage, localized heat incursion issues and potential fuel relocation/dispersal (already observed after LOCA integral tests performed on high burn-up materials).

Then, in view of being more representative of LOCA anisothermal dynamic conditions, thermal ramp tests were recently performed at CEA on 50 cm long M5 clad segments with 12-15 µm thick Cr-coating, with heating rates ranging from 0.1 up to 25°C/s and internal pressures from 10 to 100 bars. The reference HT thermomechanical testing "EDGAR" CEA facilities [START_REF] Forgeron | Experiment and modelling of advanced fuel rod behavior under LOCA conditions : α↔β phase transformation kinetics and EDGAR methodology[END_REF] have been used. Their main characteristics are recalled here after.

The EDGAR-1 and EDGAR-2 facilities were built in 1974 and 1994 respectively at CEA and present similar characteristics. These devices are devoted to characterize the thermomechanical behavior of cladding tubes during the first step of LOCA transients. Since the eighties, more than two thousand EDGAR tests have been performed at CEA on several types of cladding materials, including pre-hydrided materials as a surrogate to high Burn-Up claddings [START_REF] Portier | Influence of Long Service Exposures on the Thermal-Mechanical Behavior of Zy-4 and M5 Alloys in LOCA Conditions[END_REF]. The tests are performed on cladding tube measuring 490 mm in length and allow to apply many kinds of thermal and pressure transients, in steam environment. The specimens are heated by direct Joule effect (with a low-voltage alternative current) in order to obtain rapid increase of temperature i.e. up to 200°C/s, and internally pressurized by argon gas between 5 and 150 bars. This heating mode enables to obtain low clad thermal azimuthal gradients (less than 5°C) because it is well known that in LOCA conditions, azimuthal thermal gradients may strongly influence the circumferential elongations of the cladding up to rupture [START_REF] Erbacher | LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling[END_REF]. Thus, the primary objective of EDGAR testing was to be able to perform precise thermally controlled experiments to derive cladding (clad material dependent) HT creep constitutive laws and burst criteria, without parasitic effects. To preserve the biaxial stress state of the tube, the top grip of specimen is free of axial displacement and its weight is mechanically compensated. The temperature is measured by two methods: -a K-type thermocouple centered inside of the clad, -through quartz windows optical pyrometer measurements, operating up to 1300°C. During the test, temperature is controlled most of the time by the pyrometer but could be backcontrolled by the internal thermocouple. Additionally, some welded multi-thermocouples and, more recently, an infrared camera have been sometimes used for instrumented testing, to check the axial and/or azimuthal thermal gradients evolution upon some EDGAR test sequences.

Most of time, the experiments are made with a constant pressure measured by a differential pressure transducer. However, the EDGAR-1 facility allows increasing or decreasing the pressure over time in the range of -10 to +10 bars/s. The EDGAR-2 facility enables to increase or decrease the pressure as a function of the recorded deformation and/or temperature, or according to any given pressure versus time transient.

The cladding diameter is measured in the middle of the useful clad length during the test by a laser sheet device through the quartz window, to follow continuously the circumferential deformation of the tube and then to be able to plot and analyze the creep versus time resultant curves. The ballooning area is generally located just below the actual location of the laser clad diameter measurement. This is why, for most of the tests performed, the deformation measured continuously does not represent the maximum deformation (total elongation at rupture) but is more representative of a "uniform" clad deformation (before the deformation localization and plastic instability occurrence leading to the final rupture). All data are recorded in real time with an adjustable frequency between 1 and 10 Hz by a computer: i.e., temperature, pressure and clad circumferential deformation.

As illustrated in Figure 1, the typical post-mortem measurements are the: i) maximum circumferential deformation (At), ii) circumferential deformation at +20 mm and -20 mm of burst (Ar) 1 , iii) average circumferential deformation (εθθ,h) in the homogeneous part of the clad.

BRIEF OVERVIEW OF THERMAL RAMP (EDGAR) TESTS RESULTS

In the following, only as-received coated and uncoated M5 materials are considered. From the different ramp tests performed on 12-15µm thick Cr coated M5 clad segments, it can be observed that: -Depending on the internal pressure applied and of the heating rate, the rupture time/temperature measured are in most cases higher than the uncoated reference material ones tested at similar ramp rates. As an illustration, Figure 2 shows an example where, for the particular heating rate and internal pressure conditions applied, the time to rupture of the Cr-coated-M5 is significantly higher than the simulated 2 uncoated one. For another EDGAR test performed under 1 +/-20mm is an arbitrary axial location which has been chosen to be more or less representative of an average value between the circumferential elongation just at the burst opening ends and the one corresponding to the homogenous part of the overall (ballooned) cladding tube length. 2 The reason why the comparison is done with a simulated curve and not with an experimental one for the uncoated reference materials is to be able to take accurately into account the actual thermal-pressure ramp applied on the Cr-i i low internal pressure and more dynamic conditions, the ramp test was stopped when the Crcoated clad reached 1133°C without burst occurrence and limited creep deformation due to a security temperature threshold of the EDGAR facility. In comparison, two tests already performed in similar conditions on uncoated reference M5 clad segments induced rupture temperatures ranging between 1030 and 1090°C (βZr temperature range) so the Cr-coating seems to increase the rupture temperature by at least ~50°C in these conditions. These particular tests seem to confirm the potential HT strengthening effect of the Cr coating on the actual creep properties of Cr-coated claddings, as already observed upon isothermal creep test, especially within the 600-850°C temperature range (αZr domain). It must be mentioned that for a few cases, the data obtained on one coated clad segment is compared to two data points relying to uncoated reference M5 (already tested in the same conditions), thus increasing slightly the apparent number of data points. Additionally, some typical clad balloon and burst opening aspects are illustrated in Figure 3. It is obvious from these figures that the balloon sizes and circumferential elongations of the Cr-coated M5 are significantly lower than the uncoated M5 ones, thus confirming the tendency already observed upon isothermal creep tests [START_REF] Brachet | Behavior under LOCA conditions of Enhanced Accident Tolerant Cr-Coated Zircaloy-4 Claddings[END_REF]- [START_REF] Brachet | Behavior of Cr-coated M5™ claddings under LOCA conditions[END_REF]. It can be also observed that, for heating rates higher than or equal to 1°C/s (which are quite prototypical of most of LOCA transients), the post-mortem clad "uniform" circumferential coated clad segment which may slightly differ from the previous ramp tests performed on uncoated reference materials elongations3 are systematically lower than ~30%. While the maximum (total) elongation measured at burst location is lower than 70%. Taking into account that there is almost no significant azimuthal thermal gradient produced upon the EDGAR tests, these values can be considered as an upper (conservative) limit for the clad behavior upon LOCA, as compared to the LOCA cladding behavior in a more representative nuclear fuel sub-assembly environment (bundle effects) 4 .

As already observed on both Cr-coated Zircaloy-4 and M5 after isothermal creep testing up to rupture at different temperatures [START_REF] Brachet | Behavior under LOCA conditions of Enhanced Accident Tolerant Cr-Coated Zircaloy-4 Claddings[END_REF][START_REF] Brachet | Behavior of Cr-coated M5™ claddings under LOCA conditions[END_REF], and whatever the applied heating rate/ internal pressure values: -the chromium coating is still fully adherent with no coating spallation, after having experienced ballooning and burst, including at the vicinity of the burst where the M5 clad substrate is highly deformed. This would thus preserve the residual Cr-coating resistance capacity upon subsequent HT steam oxidation (until the final core reflooding by the Emergency Cooling Systems); -post-test surface of the coated cladding shows mainly negligible oxidation (with a typical "metallic" aspect), while the uncoated reference materials, tested in the same conditions, show a black surface due to formation of outer zirconium oxide; this confirms the good resistance of Cr coatings to HT steam oxidation (see macrographs on Figure 3); -when the rupture occurs at temperatures higher than ~850°C (i.e., corresponding to {αZr + βZr} or fully βZr temperature ranges), burst opening is generally very small, of the order or less than 1mm². This apparent intrinsic Cr-coating effect on reduction of the burst opening may have positive consequences by mitigating the potential fuel relocation and dispersal as sometimes observed after semi-integral or integral tests performed on high burnup (BU) irradiated fuel rods. Additionally, it is believed that such a small burst opening may significantly limit the steam access to the inner surface of the clad, thus delaying the potential clad embrittlement risk due to secondary hydriding phenomena.

CONCLUSIONS

Chromium coated zirconium based fuel claddings are developed and studied within the CEA-Framatome-EDF joint research program as an "Enhanced Accident Tolerant Fuel" (EATF) cladding short-term concept for nuclear light water reactors. In addition to their already demonstrated improved resistance to High Temperature (HT) steam oxidation, the present study focused on their thermomechanical behavior upon thermal ramps under internal pressure, typical of LOCA transients. When compared to uncoated reference M5 claddings tested in the same conditions, the "EDGAR" thermal ramp tests performed on 10-15 µm Cr-coated M5 confirm potential LOCA behavior improvements, preliminarily observed upon isothermal creep tests:

comparable or higher rupture temperatures (HT Cr-coating strengthening effect); -significantly lower balloon sizes and, in some conditions, extremely small burst opening sizes; -negligible HT steam oxidation and no coating spallation, even at the vicinity of the burst where the clad substrate is highly deformed, thus preserving the Cr-coating resistance capacity to subsequent HT steam oxidation.

On-going and further work -Additional LOCA ramp tests are on-going to continue to build a larger database and improve the statistics in the LOCA creep and burst behavior, to potentially derive more precise behavior models if necessary. In the future, it would be also interesting to extend internal pressure tests at HT on irradiated materials to confirm the LOCA behavior improvement observed on as-received coated materials, after achieving significant Burn-Up.

4.

Fig 1 .

 1 Fig 1. Postmortem measurements on an EDGAR tested clad segment of 490mm in length

Fig 2 .

 2 Fig 2. -Typical evolution of Cr-coated M5 clad circumferential elongation measured upon a low heating rate thermal ramp test for an intermediate internal pressure applied; comparison with the simulated uncoated M5 reference cladding behavior -In Figures 3 to 5, the post-mortem circumferential deformation parameters measured on Cr-coated M5 are compared to the uncoated reference M5 ones tested in the same conditions. It must be mentioned that for a few cases, the data obtained on one coated clad segment is compared to two data points relying to uncoated reference M5 (already tested in the same conditions), thus increasing slightly the apparent number of data points. Additionally, some typical clad balloon and burst opening aspects are illustrated in Figure3. It is obvious from these figures that the balloon sizes and circumferential elongations of the Cr-coated M5 are significantly lower than the uncoated M5 ones, thus confirming the tendency already observed upon isothermal creep tests[START_REF] Brachet | Behavior under LOCA conditions of Enhanced Accident Tolerant Cr-Coated Zircaloy-4 Claddings[END_REF]-[START_REF] Brachet | Behavior of Cr-coated M5™ claddings under LOCA conditions[END_REF]. It can be also observed that, for heating rates higher than or equal to 1°C/s (which are quite prototypical of most of LOCA transients), the post-mortem clad "uniform" circumferential

Fig 3 .Fig 4 .

 34 Fig 3. Maximum circumferential deformations (At) for Cr coated M5 compared to those measured for uncoated references, tested in the same conditions (Temperature range: 0.1°C/s to 25°C/s, Pressure range: 10 to 100 bars)

Values measured at +/-20 mm from the burst opening and averaging values for the homogenous part of the cladding, which correspond to Figures

and

respectively[START_REF] Forgeron | Experiment and modelling of advanced fuel rod behavior under LOCA conditions : α↔β phase transformation kinetics and EDGAR methodology[END_REF] This last situation inducing likely some azimuthal thermal gradients due to the thermal influence of the neighboring nuclear fuel rods, leading to a decrease of the cladding total elongation at rupture, as discussed previously


M Gherrab
email: mehdi.gherrab@cea.fr

F Dacquait

R Lecocq

ENGIE K Schildermans

Influence of the dissolved hydrogen concentration on the contamination of the primary loop of DOEL PWR using the OSCAR code

. OSCAR is a reliable tool for PWR (also used for EPR, SFR, ITER, decommissioning…) calibrated and validated with a complete database of contamination measurements in EDF water reactors.

Water chemistry has an influence on corrosion of the main materials (especially nickel-based alloys), in Belgian PWR the average dihydrogen concentration used is around 30 cc/kg, which is in line with current international standards however which is not the best value to reduce stress corrosion cracking of the materials. It also has an influence on dissolution/precipitation mechanisms involved in contamination. Water chemistry control allows then to reduce significantly the radioactive contamination in the primary loop and therefore facilitates maintenance operations. In this field, hydrogen plays a critical role in limiting the presence of oxidizing species due to water radiolysis. Increasing hydrogen could also reduce core internals cracking. The aim of this study is to evaluate the influence of an increase/decrease of the Dissolved Hydrogen (DH) concentration on the contamination of the primary loop using the OSCAR code. This study presents the simulation results of a sensitivity analysis, using the 1.3 version of the OSCAR code, of the contamination in the primary loop of DOEL PWR with DH concentrations ranging between 15 and 70 mL/kg.

2

The evolution of the surface contamination in 58 Co and 60 Co were calculated on the hot legs, U tubes and Steam Generators (SG) tubing along with the mass of Ni deposited on the fuel. In order to explain those evolutions, equilibrium (concentration of the element in the coolant with respect to the considered oxide) and ionic Ni concentrations were investigated on the SG and the fuel as well as the Ni dissolution and release flux of the SG.

Introduction

Normal operation of PWR generates corrosion products in the primary circuit, which are activated in the core and constitute the major source of the radiation field. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products by carrying out the appropriate measurements in PWR circuits and loop experiments combined with reactor contamination assessment code. The reactor contamination assessment code allows to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. The OSCAR code [START_REF] Dacquait | The OSCAR V1.4 code: a new dissolution/precipitation model to better simulate the corrosion product transfer in nuclear cooling systems[END_REF] has been developed with this aim by CEA in collaboration with EDF and Framatome, and has actually been used since the early seventies. OSCAR is a reliable tool for PWR (also used for EPR, SFR, ITER, decommissioning…) calibrated and validated with a complete database of contamination measurements in EDF water reactors. Water chemistry has an influence on corrosion of the main materials (especially nickel-based alloys), in the Belgian PWRs the average dihydrogen concentration used is around 30 cc/kg, which is in line with current international standards however which is not the best value to reduce stress corrosion cracking of the materials. It also has an influence on dissolution/precipitation mechanisms involved in contamination. Water chemistry control then allows to reduce significantly the radioactive contamination in the primary loop and therefore facilitates maintenance operations. In this field, hydrogen plays a critical role in limiting the presence of oxidizing species due to water radiolysis. Increasing hydrogen could also reduce core internals cracking. The aim of this study is to evaluate the influence of the change in the Dissolved Hydrogen (DH) concentration on the contamination of the primary loop using the OSCAR code. This study presents the simulation results of a sensitivity analysis, using the 1.3 version of the OSCAR code, of the contamination in the primary loop of DOEL PWR with DH concentrations ranging between 15 and 70 mL/kg. It also presents autoclave experiments to evaluate the impact of several DH concentrations on INCOLOY 690 material, which are compared to simulation results.

Operating Parameters of DOEL-4

DOEL-4 is a 3-loop PWR equipped with INCOLOY 690 tubed steam generators since the steam generator replacement at cycle 11. The structure of the DOEL-4 primary circuit has been modelled by control volumes using the design data of the reactor (wetted surfaces, hydraulic diameters, material compositions, nominal temperatures …). Operating cycles of DOEL-4 were simulated using the 1.3 version of the OSCAR code with the real hydrogen concentrations for cycles 1 to 22. Three more reference cycles were added to DOEL-4 to try various hydrogen concentrations [START_REF] Schildermans | OSCAR Influence of RCS hydrogen concentration on surface contamination[END_REF]. The parameters of the reference cycles are reported in Table 1. The operating parameters (Power, CB, CLi, CH2, CO2) are given in Figure 1.

The last three cycles correspond to the reference cycles with a simulated shutdown at the end of each cycle.

Power is set to 100%, the boron concentration decreases at each cycle from 1424 pm to 157 ppm, the lithium concentration decreases from 3.35 ppm to 0.59 ppm. Concerning hydrogen, three cases are studied 15, 30 and 70 mL/kg with a stable value during the three cycles for each case. The oxygen peaks observed are due to the shutdowns. Concerning 58 Co, it is clear that an increase of the DH value leads to a decrease in the surface activities in the primary system on the last three cycles of DOEL-4. Concerning 60 Co, the decrease tendency is slight.

For DH concentrations of 15 and 70 mL/kg, the relative variations of the total activities (of 58 Co and 60 Co) deposited on the out of core surfaces at the end of cycle 25 compared to a DH value of 30 mL/kg are presented in Table 2.

The increase in the DH concentration up to 70 mL/kg leads to a decrease in the total activities of 58 Co and 60 Co respectively by 57% and 4%.

On the contrary, a decrease in the DH down to 15 mL/kg leads to an increase of the total activities of 58 Co and 60 Co respectively by 74% and 4%. It also shows that an increase in the DH concentration (from 15 to 70 mL/kg) lead to a decrease of the total activities of 58 Co and 60 Co.

Ni and Fe masses deposited on the fuel rods are presented in Figure 6. The Ni mass deposited on the fuel decreases when the DH concentration increases (same evolution as the surface activities). On the contrary the Fe mass deposited on the fuel increases when the DH concentration increases.

DUPLEX experiments

The DUPLEX experimental device [START_REF] You | CEA internal report[END_REF] is composed of two titanium (TA6V) columns (see Those experiments confirmed the OSCAR results obtained in Figure 6.

Figure 1 :

 1 Figure 1: Operating parameters

Figure 2 :Figure 3 :

 23 Figure 2: Surface activities on the Hot Leg

Figure 4 :

 4 Figure 4: Surface activities on the hot side of the Steam Generator tubing

Figure 5 :

 5 Figure 5: Surface activities on the cold side of the Steam Generator tubing

Figure 6 :

 6 Figure 6: Mass deposited on the fuel

Figure 7 :

 7 Figure 7: Equilibrium and ionic Ni concentrations in the fuel region

Figure 8 :

 8 Figure 8: Equilibrium and ionic Ni concentration on the cold side of the steam generator

Figure 10 :

 10 Figure 10: Ni dissolution versus corrosion release rate in the steam generator for the cycle 25

Figure 11 :

 11 Figure 11: Co dissolution versus corrosion release rate from the steam generator tubing for the cycle 25

  Figure 12) crossed by a steady-state and monophasic flow. This device allows testing various DH concentrations. The bottom column, set with a temperature of 270 °C, reproduces the steam generator conditions and contains alloy 690 shavings. The upper column, set with a temperature of 340 °C, reproduces the fuel cladding and contains preoxidized M5 alloy. This upper part is only dedicated to the study of the precipitation of the corrosion products coming from the bottom part. The elementary composition, especially Ni composition, of different solutions is obtained by mass spectrometry.

Figure 12 :

 12 Figure 12: DUPLEX device

Figure 13 :

 13 Figure 13: Ni mass released from alloy 690 for various DH values

Table 1 :

 1 Parameters of the reference cycles

	Nominal	Cycle	Binit	Liinit	DH	pH
	Power	duration
	100%	473 days	1424 ppm	3.35 ppm	15 -30 -70	7.2 (at 312 °C)
					mL/kg

Table 2 :

 2 Relative variations of the total activities deposited on the out of core surfaces compared to a DH value of 30 mL/kg

	Variations /	Out of core surface activities
	[H2] = 30 mL/kg Cycle 25	58 Co	60 Co
	[H2] =15 mL/kg	+74%	+4%
	[H2] =70 mL/kg	-57%	-4%

This table shows that DH concentration significantly affects the 58 Co contamination of the out of core surfaces and to a negligible extent the 60 Co contamination.

Table 3 :

 3 Ni and Co ratios at the end of cycle 25This table shows that DH concentration significantly affects the Ni rate in the SG, which impacts, due to neutron activation in the fuel regions, the 58 Co contamination of the out of core surfaces.

	Ratios /	Ni rate in the SG	Co rate in the SG
	[H2] = 30 mL/kg Cycle 25	Dissolution	Corrosion release	Dissolution	Corrosion release
	[H2] =15 mL/kg	2.3	1.3	0.2	1.3
	[H2] =70 mL/kg	0.3	0.8	4.0	0.8

Concerning Co.

Equilibrium and ionic Co concentrations in the fuel regions and the steam generator in the case of a DH value of 70 mL/kg are presented in Figure 9. Equilibrium Co concentrations are above ionic Co concentrations in the fuel and the steam generator. As a result, Co tends to dissolve and slightly precipitate. The peaks observed on the last three cycles correspond to the simulated shutdowns.


This immense richness was a key element for the maintenance of the position of supreme power (through public expenditures or benefactions on individuals and communities), and had an important economic value.

No systematic survey of all the available documentation has been produced since the beginning of the 20th century; many questions about their acquisition and use remain.

Patrimonivm will take an empirical, bottom-up approach to imperial properties, starting from the relevant sources that are collected in a single, online database, the Atlas patrimonii Caesaris, in order to draw more general and theoretical conclusions.

Visualize distribution of properties & people

Map economic connections & human mobility yanne.broux@u-bordeaux-montaigne.fr or yanne.broux@kuleuven.be For more information about the Patrimonivm project in general, please visit: https://patrimonium.huma-num.fr/project.html or contact Alberto Dalla Rosa (PI): alberto.dallarosa@u-bordeaux-montaigne.fr

KEY RESEARCH AXES

• Geographical distribution: variations in size and importance per region.

• Patterns of growth: mainly passively through inheritances, donations, or confiscations ➞ understand local distribution of private property!

• Economic value: different factors, e.g. profit for the fiscus, size of nearby available resource-consuming centers, accessibility of markets.

• Administration: regional and sub-regional divisions in relation to size of properties and relevance for most (local) markets.

• Interaction center-periphery: interaction with Roman administration and law + mobility of administrators and people working on properties.

Updated overview of sources

Key topics:

• imperial vs "non-imperial" οὐσίαι (Parássoglou, Imperial Estates in Roman Egypt, 1978)

• sale of imperial property

• regional differentiation in management and exploitation (Fayum vs Middle Egypt)

• the ousiakos logos and the procurator usiacus

  for the study of imperial properties will be collected in a single relational XML database, consisting of four main tables: • Sources: epigraphic, papyrological, literary & archeological • People: owners, administrative personnel, tenants, … • Places: geolocation of properties • Bibliography: editions, secondary literature, websites, … Imperial properties are attested in all corners of the Roman Empire and included residences, cultivated land, pastures, woods, mines, quarries, luxury items and slaves.


Mona P Moret
email: mona.moret@ametek.com

Claire Le Gall

Karine Hanifi

Isabelle Felines

Thierry Blay

Anne-Sophie Robbes

Carl Henderson

Fabienne Audubert

Yves Pontillon

Jerôme Lamontagne

Shielded Electron Microprobe and some of its main applications in Hotlabs

Keywords: EPMA, shielded electron microprobe, quantitative analysis, nuclear fuel, fission products

à la diffusion de documents scientifiques de niveau recherche, publiés ou non, émanant des établissements d'enseignement et de recherche

Abstract: Thanks to its precision, its reproducibility and its stability, Electron Microprobe is a well suited technique for accurately analyzing nearly all chemical elements at concentration levels down to few 10's ppm with a spatial resolution of about 1 µm, which is relevant to microstructures in a wide variety of materials and mineral specimens. For irradiated samples, EPMA is also one of the technique of choices and can support nuclear fuel development, control, metallurgical or glass analysis. It reveals fine compositional details and the distribution of main and trace elements across the surface of the sample. CAMECA leader in scientific instruments has been manufacturing Electron Microprobe (EPMA) since 1958 and will present its latest CAMECA shielded EPMA model, SKAPHIA released in 2016 for hotlab facilities. One example of the LECA/STAR shielded EPMA in hotlab handling nuclear fuel will be highlighted. In this example, the microstructure and chemical evolution of fuel pellets submitted to two different nuclear severe accident scenarii are shown. EPMA helped to highlight the impact of the atmosphere (i.e the oxygen potential) on the behavior of the fuel and fission products during these tests. These studies are of particular importance to evaluate and better predict the consequences of such an accident in term of contamination.

Shielded

  Electron Microprobe and some of its main applications in HotlabsCharacterization of MOX irradiated fuels treated at very high temperatures A. Moret, C. Le Gall, K. Hanifi, I. Felines, T. Blay, A.-S. Robbes, C. Henderson, F. Audubert, Yves Pontillon, J. Lamontagne To cite this version: A. Moret, C. Le Gall, K. Hanifi, I. Felines, T. Blay, et al.. Shielded Electron Microprobe and some of its main applications in HotlabsCharacterization of MOX irradiated fuels treated at very high temperatures. HOTLAB, Sep 2018, Helsinki, Finland. cea-02328971


J B Génin

T Jobert

The OSCAR-FP V1.4 code: Simulation of fission product and alpha emitter contamination in PWR circuits

Understanding the PWR primary circuit contamination is a crucial issue for reactor operation and design, as well as in the perspective of decommissioning. For this purpose, the CEA has developed, in cooperation with EDF and Framatome, the OSCAR code package, a unique tool to simulate PWR contamination by corrosion products, fission products (FP), and actinides. This paper presents a general overview of OSCAR-FP V1.4, the latest version of OSCAR computation scheme dealing with fission product and actinide contamination. The source term calculation for these nuclides and the behavior considered in the primary circuit are described in detail, as well as the validation process against primary activity measurements.

Fission products are released in the primary circuit in case of a fuel cladding defect, depending on the nuclear fuel characteristics: composition, temperature, and burn-up. Therefore, the CEA fuel code ALCYONE is included in OSCAR-FP and allows calculating the source term for these nuclides. However, fission product release into rod free volumes is increased due to fuel oxidation in case of a clad defect. The fuel code ALCYONE has then been adapted to defective fuel situation through the adjustment of different parameters: enhanced diffusion in the fuel grain, enhanced grain boundary fracturing, and reduced fuel thermal conductivity.

In addition, in the event of significant cladding damage, fuel oxide particles may disseminate in the primary circuit. These particles solubilize very slowly and are still present in the primary circuit several cycles after the dissemination. They include long life fission products as well as heavy nuclides among which alpha emitters and fissile nuclides. They are mainly deposited on the primary surfaces and are subject to further erosion and deposition mechanisms. As their behavior is linked to the one of corrosion product particles, a specific OSCAR computation scheme is used, allowing coupled simulation between corrosion products and actinides. Calculated primary activities are compared to measured fission product and alpha activities, obtained from primary gamma activity monitoring as well as from sampling and analysis campaigns of primary coolant or surface deposits. Code qualification is obtained for different types of nuclides: i) gaseous and volatile gamma emitters Xe, K, I, Cs whatever the isotope half-life, ii) alpha emitters, which are mainly insoluble, iii) long life beta emitter 90 Sr, which undergoes adsorption/desorption mechanism and is sensitive to pH.

The paper shows that the OSCAR-FP V1.4 code is a powerful tool to study fission product and alpha contamination in PWR primary circuits, and is helpful for fuel and waste management as well as for the alpha risk assessment. Moreover, new results obtained with the V1.4 version are highlighted: mechanistic erosion and deposition models have been calibrated for corrosion products and applied to actinides. This provides a good simulation of actinides with very few adjusted parameters, and confirms the coupled behavior of corrosion products and actinides in the primary circuit.

Introduction

Different radioactive nuclides are responsible for the radioactive contamination of the PWR primary circuit. They mainly come from bulk corrosion of primary materials (stainless steel, nickel alloy, stellites…), which induces the release of several kilograms per year of corrosion products in the primary circuit. These corrosion products become radioactive when exposed to neutron flux. They are the main contributor to the dose rate around the primary circuit. However, in case of defective fuel cladding, radioactive fission products also contaminate the primary circuit. In some rare cases, a few grams of fissile material including alpha emitter actinides may also be released in the primary circuit.

While operating or designing a nuclear power plant, it is necessary to address important challenges related to the circuit contamination: optimizing the personnel exposure to radiation, limiting the activity of the wastes and effluents produced during the reactor lifetime and preparing decommissioning. Therefore, the assessment of the radio contamination of PWR circuits is a crucial issue for power plant operators and suppliers.

To address this issue, the OSCAR code package is targeted at simulating PWR primary and auxiliary circuit contamination in nominal and cold shutdown conditions. It is qualified for corrosion product, fission product and actinide contamination. This paper focuses on the simulation of fission products and actinides with the latest OSCAR version (V1.4). However, as corrosion products represent most of the contamination mass deposited on the primary surfaces, they definitely determine the evolution of other deposited radionuclides, in particular actinides. Thus, the coupled behavior between corrosion products and actinides is considered.

After a general overview of the code basics in section 2, the paper presents the main principles of modeling as well as the validation process for fission products and actinides in case of defective fuel cladding (section 3) and in case of fissile material dissemination (section 4).

General overview

The OSCAR code package has been developed within the framework of R&D cooperation between the French Atomic and Alternative Energy Commission (CEA), Electricité de France (EDF) and Framatome. The code uses a fully object-oriented design and a modular structure. Circuits are discretized into as many regions as necessary, according to the geometric, thermal-hydraulic, neutronic, material and operating characteristics. A user-friendly interface for input data and visualisation of results is included.

Even though the origin is different for corrosion products, fission products, and actinides, they are subject to the same severe conditions (300 °C, 150 bar, neutron fl ux, water velocity up to 15 m.s -1 ) and follow the same transport mechanisms in the primary circuit. The main processes involved are erosion/deposition, dissolution/precipitation, adsorption/desorption, convection, purification, neutron activation, radioactive decrease [START_REF] Génin | The OSCAR code package : a unique tool for simulating PWR contamination[END_REF]. The different mechanisms can be activated separately. Taking into account these mechanisms, mass balances are calculated for each region and for each relevant nuclide (radioactive and stable), either at the metal surface or in the primary coolant (particles and soluble forms). Regarding fission products, 78 nuclides within 27 filiation chains are handled in the OSCAR code. For actinides, a 22 nuclide filiation chain is considered.

Three different computation schemes are now proposed to the OSCAR user. The two original schemes (OSCAR-CP and OSCAR-FP) are dedicated to corrosion products and to fission products and actinides respectively. They can be run independently. However, considering the coupled behavior of corrosion products and actinides, a third scheme (OSCAR-CPFP) has been developed. It allows calculating all types of radio contaminant with the same data set. In return, the data set is more complex as it gathers the data needed for the simulation of the different elements. Compared to the original schemes, the third scheme has no impact on corrosion product simulation as actinides and fission products do not affect the corrosion product behavior. It is useful in case of fissile material dissemination (see § 4).

Fission product release out of defective rods

In the event of a rod cladding defect, gaseous (Xe, Kr) and volatile (I, Cs) fission products are released into the primary circuit. The process involves several stages: release out of the rod pellets into the free volumes, migration along the fuel-cladding gap and release through the defect into the primary coolant [START_REF] Génin | The OSCAR code package : a unique tool for simulating PWR contamination[END_REF].

Once in the primary circuit, the behavior of these species is simply modeled:

• They are considered as fully soluble.

• They do not interact with the primary walls in the nominal operating conditions.

• The Henry's law coefficient is used to evaluate the partition between liquid and gaseous phases (purification tank, pressurizer). • Volatile fission products are trapped on the ion exchange resins, whereas gaseous species may be removed through gas flush.

The concentration of the gaseous and volatile fission products in the primary coolant is then easily calculated if the release rate is known. As far as these species are concerned, the assessment of the release rate from a defective rod is the main challenge for the OSCAR code. The other species are supposed not to be released because their diffusion in the fuel matrix is not fast enough.

Release out of rod pellets

Because of the presence of water or vapor inside the defective rod, oxidation of the fuel matrix occurs, which strongly enhances the release of fission products out of the rod pellet compared to a tight fuel rod. Modeling of this phenomenon should rely on defective fuel characterization but given the lack of such experiments, the strategy has been to adapt the MARGARET module developed and validated for tight fuel.

MARGARET [START_REF] Noirot | MARGARET: A comprehensive code for the description of fission gas behavior[END_REF] is an advanced mechanistic model of fission gas behavior in PWR fuel. It provides the detailed location of the gases -inter-or intragranular, in bubbles or dissolved -, in addition to gas release and swelling. It is developed by the CEA within the framework of the PLEIADES platform [START_REF] Michel | Simulation of pellet-cladding interaction with the PLEIADES fuel performance software environment[END_REF] and its fuel behavior application ALCYONE, which supplies the thermomechanical data used by MARGARET.

The MARGARET code, which initially dealt only with stable gases, has been adapted to the radioactive fission products. At each time step, a first calculation is performed for stable gases and provides the fuel microstructure: gas bubble distribution as well as grain boundary fracturing or fuel restructuring into HBS (High Burnup Structure). Then a second calculation is performed for radioactive fission products, using the previously calculated fuel microstructure and taking into account the radioactive decay constant and the filiation chain.

Like stable gases, radioactive fission products are released from the grain by diffusion, either towards grain boundaries or towards free surfaces, that means grain surfaces connected with free volumes (fracture or open porosity or fuel-cladding gap).

When released at free surfaces, fission products are considered to migrate instantaneously to the free volumes. When released at the grain boundary, they may migrate towards the free volumes by percolation, when the bubbles at the grain boundaries interconnect at high burnup. The free surface fraction is simulated by the MARGARET code, and is around 1 or 2 % in a tight fuel configuration.

Moreover, to fit the experimental release of short life radionuclides at low temperature, the knockout mechanism [START_REF] Olander | Fundamental aspects of nuclear reactor fuel elements[END_REF] at the grain free surface has been added. Knockout means ejection of an atom struck by a fission fragment or another atom of the lattice, due to elastic collisions. The important parameter for this mechanism is the thickness of the layer where knockout can lead to the ejection of fission products out of the grain. This thickness is assessed about 15 nm. The knockout flux can be expressed as an additional sink term in the knockout layer to the diffusion term. This flux is proportional to the fission rate and to the fission product concentration in the layer. Knockout does not significantly affect released fractions by diffusion for 133 Xe (5 day period) and 131 I (8 day period), which are used for defective fuel calibration (see § 3.3). However, it strongly increases the release of shorter half-life fission products, particularly at low fuel temperatures, as they do not have time to diffuse out of the grain [START_REF] Génin | The OSCAR code package : a unique tool for simulating PWR contamination[END_REF].

Thus, mechanisms involved in fission product release into the free volumes are mainly: diffusion within the fuel grain, knockout at the free surface of the grain, and percolation through the grain boundary at high burnup. The behavior is considered to be the same for the different species (Xe, Kr, I, Cs). In particular, the intragranular diffusion coefficient has the same value for each of these species. Only the radioactive decay period and the filiation affect the released fraction. In future, a chemical module will be included in MARGARET, taking into account the solubility of elements in the fuel matrix.

In the event of a cladding defect, most of the primary fluid activity is due to xenon, krypton and iodine isotopes with radioactive half-life shorter than 8 days (see Figure 1). The activity of these radioactive nuclides is continuously monitored during reactor operation.

If the burnup is not too high (no percolation) and the irradiation power not too low (knockout is negligible), the release of these isotopes of interest is due only to diffusion out of the grain. In first approximation, the diffusion coefficient is an apparent diffusion coefficient, taking into account not only the intrinsic intragranular diffusion coefficient but also the trapping and re-solution coefficients of pores and bubbles. This approximation is always valid for long life fission products and for shorter life nuclides as long as precipitation bubbles do not appear in the grain (burnup lower than 30 GWd/t).

The released fraction can then be expressed as:

3 pellet app free R D F B a λ = ⋅ ⋅ (1)
where pellet R is the release rate from the pellets into the free volumes, B is the fission product birth rate in the pellets, free F is the fraction of grain surfaces connected with free volumes, a is the grain radius, app D is the apparent diffusion coefficient of the fission product in the grain, and λ the radioactive decay constant of the fission product.

In the logarithmic diagram of the released fractions out of pellets versus the radioactive decay constant, the maximal absolute value of the slope is ½ if the burnup is lower than 30 GWd/t. The slope is lower at low power because knockout becomes faster than diffusion for short life fission products.

Release out of the rod

Once released from the pellet, as long as the fuel-cladding gap is not closed, noble gases are supposed to migrate to the defect rapidly, where they are released into the primary coolant according to the conductance of the defect. However, at high burnup, the gap is partially or totally closed and the migration is far slower.

Considering iodine and cesium isotopes, the release mechanisms are not well understood. During nominal reactor operation, trapping occurs in free volumes, which makes the release rate out of the rod very difficult to assess. However, during power transients at low level, water enters the free volumes and leads to the release of almost the total amount of iodine and cesium present in free volumes.

This behavior is taken into account in the modeling through two parameters:

• The conductance of the defect: ν (s -1 )

• The trapping coefficient in the free volumes (ratio of trapped to non-trapped activity): α

According to the modeling, the rod release is the product of defect conductance and non-trapped activity in free volumes. Then, the activity balance in free volumes is the following [ A (Bq): total activity of the fission product in the free volumes, pellet R (Bq/s): release rate from the rod pellets into the free volumes, rod R (Bq/s): release rate from the rod into the primary coolant, λ (s -1 ): radioactive decay constant]:

pellet rod dA R A R dt λ = -⋅ - , with 1 rod R A ν α = ⋅ +
In steady state, the relationship between the instantaneous released fractions from rod and pellets is:

1 1 pellet rod R R B B ν α ν λ α + = ⋅ + +
This model is very simple and cannot account for high burnup defective rod, where gases are mechanically retained.

Moreover, the parameters ν and α are difficult to assess, though ν can be related to the defect size.

Nevertheless, the model is useful for interpreting the instantaneous released fractions in nominal operation ( ), rod R B which are obtained from the monitored primary activities and the calculation of the birth rate B .

For noble gases,

0 α = if the fuel-cladding gap is not closed. Then pellet rod R R B B ν λ ν = ⋅ + Moreover, if λ ν
<< (this is generally the case for the monitored radionuclides if the size defect is not too small), we obtain

pellet rod R R B B =
: the instantaneous released fractions from the pellets and from the rod are equivalent.

For volatile species, the trapping coefficients are assessed to 600 and 10 000 in nominal operation, respectively for iodine and cesium.

Then

1 ν λ α >> +
for the nuclides of interest.

In Figure 1, the logarithmic diagram shows the simulated rod released fractions in nominal operation versus the radioactive decay constant of the fission products. It represents the ideal case where the burnup is low, the size defect is large and the irradiation power is high enough to neglect the knockout mechanism. In this case, the theoretical slope is around 1 2 for noble gas isotopes (red line) and 3 2 for iodine isotopes (green line). The blue line represents the released fractions in case of fissile material contamination and will be described in section 4.

Figure 1 -Fission product released fractions in nominal operation

Lastly, the MARGARET code is used to assess the iodine and cesium activity which is trapped in free volumes. This activity is released during a power transient and leads to an activity peak in primary coolant. Moreover, the code allows setting to zero the trapping coefficient during the transient and then calculating a maximum peak height considering that iodine and cesium are released according to the defect conductance. However, it cannot reproduce the actual kinetic of the iodine or cesium release which is observed during the transient and is often slower than the theoretical one.

Validation method

The ALCYONE/MARGARET code is validated for tight fuel. The latest version has been adapted to simulate the release from the pellets in case of a cladding defect.

For defective rods, oxidation of the fuel matrix leads to released fractions much higher (by a factor of roughly 30) than for non-defective fuel rods. This impact from oxidation is taken into account in the simulation by the following adjustments:

• Fraction of grain surfaces connected with free volumes free F : it is evaluated to 30%.

• Intragranular diffusion coefficient D : a correction factor is applied to the intrinsic stable gas diffusion coefficient used for tight fuel. It is assessed to 1.1.

• Fuel temperature: besides the fact that for a defective configuration, water vapor replaces helium in the free volumes of the rod and that the internal pressure of the rod is balanced with external pressure (no further clad creeping), a degradation of the thermal conductivity of the fuel is taken into account (factor 0.65).

The calibration of these model parameters (fraction of free surface, intragranular diffusion coefficient and thermal conductivity degradation) has been performed against a unique experimental database including defective PWR rods well identified and characterized, as well as analytical irradiations in the SILOE reactor.

Actual fuel released fractions are obtained as follows from the primary activity monitoring:

• The 133 Xe released fraction from the fuel in nominal operation is supposed to be equal to that of the rod, as 133 Xe is the longest half-life noble gas which is monitored (see Figure 1). Then it is not affected by the transfer in the fuel-cladding gap or by the size of the defect. However, the absence of peak during power transients must be checked, ensuring that gases are not mechanically retained in the rod in nominal operation. • The 131 I released fraction from the pellets is obtained from activity monitoring during transients, considering that activity trapped in the free volumes is totally released during the transient.

As they are likely not affected by the knockout mechanism, the fuel released fractions obtained for both radionuclides correspond to equation [START_REF] Génin | The OSCAR code package : a unique tool for simulating PWR contamination[END_REF]. They should be at the same level as their radioactive decay constant is similar. This allows reducing the uncertainty about the measured released fraction.

Figure 2 gives a comparison between measurements and OSCAR calculations after calibration, concerning the released fractions from defective UO2 fuel rods of the database (including EDITH 02 and EDITH 06 analytical irradiations, respectively for low burnup and 50 GWd/t UO2 rods). The validation domain is up to 50 GWd/t. We note a relatively good agreement (within a factor of 2) between experimental and calculated released fractions. Such a fair agreement (not shown here) is also obtained for analytical irradiations EDITHMOX 01 (low burnup MOX rod) and EDITHMOX 02 (50 GWd/t MOX rod), which allows widening the validation domain to MOX fuel, for the same burnup range.

From the simulation of the defective rods of the experimental database, we observe that the burnup activated mechanisms are negligible for 133 Xe and 131 I: intergranular percolation, trapping in intragranular precipitation bubbles, HBS restructuration. Nevertheless, these mechanisms could become significant at higher irradiation power.

Validation domain

Validation of released fractions out of the fuel pellets is obtained for 133 Xe and 131 I. However, the validation domain can be extended to shorter and longer half-life nuclides.

For iodine and cesium isotopes, the fuel released fractions are obtained from activity monitoring during power transient. The integral of the primary activity peak corresponds to the amount trapped in free volumes before the transient. It is related to the instantaneous release (Bq/s) for short life nuclides ( 131 I, 133 I) and to the cumulated release (Bq) for long life nuclides ( 134 Cs, 137 Cs). Indeed, steady state is not reached for long life nuclides. Uncertainty on cesium release obtained from primary activity measurements is higher than for iodine because of the unknown operation time for cationic resins. Moreover, the uncertainty is higher for 133 I than for 131 I due to the shorter radioactive period. However, despite the higher uncertainty, the measured released amounts for 134 Cs (period: 2 years), 137 Cs (period: 30 years) and 133 I (period: 21 hours) are consistent with the calculated ones, allowing the extension of the validation domain to each iodine and cesium isotope.

Regarding noble gas nuclides, the shortest half-life nuclides are sensitive to the defect conductance, leading to a steeper slope for the released fractions in the primary circuit versus the radioactive decay constant than for the pellet released fractions. We can verify that measured released fractions in the primary coolant are lower than or equal to the calculated pellet released fraction, and that they can be simulated adjusting the defect conductance if the fuel-cladding gap is not closed.

Fig. 2 -Comparison between experimental (black symbols) and calculated (red symbols) fuel released fractions

Finally, the validation domain for gaseous and volatile fission product release in the OSCAR V1.4 code is the following:

• Fuel:

o UO2 and MOX, o Burnup range up to 50 GWd/t, o Irradiation power range up to 220 W/cm. • Fuel released fraction: validated for every isotope of Xe, Kr, I, Cs. The released fraction is sensitive only to the radioactive period. Difference between calculation and measurement is lower than a factor of 2 in the validation domain (see Figure 2).

• Released fraction into the primary circuit during nominal operation: validated for every gaseous isotope (Xe, Kr) if the conductance of the defect can be assessed and if the gap is not closed. A conservative value can be obtained considering a very high defect conductance.

• Total amount of activity released during a power transient: for every volatile isotope (I, Cs), a conservative value can be obtained considering no release during nominal operation. Then the total released amount corresponds to the calculated activity in the free volumes before the power transient.

Defective fuel diagnosis methods

Diagnosis methods of the defective fuel characteristics have been developed and gathered in the DIADEME software. These methods are based on the previously described modeling and have already been presented in detail [START_REF] Génin | DIADEME: a computer code to assess in operation defective fuel characteristics and primary circuit contamination[END_REF]. They are used along with continuous gamma primary activity monitoring and are helpful for the operator to assess in operation the number of failed rods and their characteristics and to prepare the failed fuel replacement for the next core loading.

The guidelines of the methods are the following:

• Mean defective rod power: related to the slope of fuel released fractions versus radioactive decay constant.

The absolute value of the slope increases with the irradiation power because diffusion becomes faster than knockout.

• Number of defective rods: related to the ratio of measured 133 Xe fuel released fraction to the calculated released fraction at the mean defective rod power.

• Defect size: related to the ratio of 131 I to 133 Xe released fractions into the primary coolant in operation. The ratio increases with the defect size.

• Defective rod burnup: related to the 134 Cs/ 137 Cs activity ratio measured in the primary coolant during power transients. The ratio increases with the defective fuel burnup.

• Fuel nature (UO2 or MOX) of the defective rod: related to the 85m Kr/ 135 Xe activity ratio measured in the primary coolant. The ratio is higher for MOX fuel.

Fuel oxide dissemination

In addition to fission product release, fuel oxide may disseminate into the PWR primary circuit in the event of significant cladding damage. The disseminated fuel oxide is also called "tramp uranium": though mainly deposited on the primary surfaces, it is continuously transferred between the different parts of the circuit through the erosion and deposition mechanisms.

The disseminated fuel oxide consists of actinides and fission products, which contaminate the primary circuit. Actinides include alpha emitters, which induce alpha contamination, and fissile nuclides (mainly 235 U and 239 Pu), which generate fission products directly in the primary circuit when under neutron flux. Fuel oxide dissemination causes a long-term contamination. Indeed, trapping on purification filters is not efficient as the actinide fraction transported in the primary coolant is very low.

Actinide contamination behavior

Fissile material disseminates in particulate form and is mainly insoluble in the primary circuit. The ALCYONE code is used to give the particle composition, which corresponds to the one of the disseminating fuel at the pellet surface.

Fuel oxide particles are mainly deposited on the primary surfaces. However, they undergo erosion and deposition, which induce transfer between out-of-flux surfaces and under flux surfaces. Consequently, fuel assemblies reloaded after the dissemination become rapidly contaminated with actinides.

When some fissile material is deposited under neutron flux, fission products are emitted directly by recoil in the primary coolant. As the recoil mechanism is immediate, the fission product released fractions are independent of the radioactive period. The blue line in Figure 1 illustrates this behavior for a few grams of disseminated material. Moreover, due to its short radioactive period (54 minutes), 134 I is the tracer of the fissile material under neutron flux. Indeed, as shown in Figure 1, its release from a defective rod is negligible: it can be measured in the primary coolant only in case of fissile material contamination.

Lastly, uranium oxide may solubilize. According to OSCAR modeling, it induces the separation between uranium and transuranium elements in the primary circuit. Then each species evolves separately, but transuranium elements remain in particulate form in nominal operation [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF]. The rate of uranium oxide dissolution is still not well known. It seems to be very low in nominal operation but significant in oxidizing conditions. In any case, actinides in the primary circuit likely remain in the form of the disseminated fuel oxide during many cycles.

Modeling improvements for actinide simulation in OSCAR latest version

In the OSCAR previous version, empirical deposition and erosion rates were specified in the data set for actinides, except during cycles following the dissemination where corrosion product deposition rates were used [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF]. With the last version V1.4, according to the hypothesis that actinides are transported by corrosion product particles, the goal is to use for actinides the erosion and deposition models which have been validated for corrosion products.

The deposition and erosion rates for corrosion products are obtained from mechanistic models [START_REF] Dacquait | Simulations of corrosion product transfer with the OSCAR V1.2 code[END_REF][START_REF] Dacquait | The OSCAR V1.4 code: a new dissolution/precipitation model to better simulate the corrosion product transfer in nuclear cooling systems[END_REF]. The particle deposition rate takes into account 1) the turbulent diffusion and the effects of inertia, 2) the sedimentation for horizontal ducts and 3) the thermophoresis for temperature gradients between the coolant and the wall. Regarding the erosion rate, it results from the coolant friction forces, and involves the deposit thickness. These mechanisms have been calibrated for corrosion product simulation [START_REF] Dacquait | The OSCAR V1.4 code: a new dissolution/precipitation model to better simulate the corrosion product transfer in nuclear cooling systems[END_REF].

Before actinide calculation, a preliminary corrosion product calculation must be performed to evaluate the corrosion product deposited mass, which is needed for the erosion mechanism. Therefore, the OSCAR-CPFP computation scheme is used as it allows calculating corrosion products as well as actinides and fission products with the same data set. Consequently, it is possible to use the same deposition and erosion mechanisms for every insoluble species.

However, compared to corrosion products, the erosion mechanism is adapted for actinides as follows:

• The erosion flux for each isotope is proportional to the occupied surface fraction m is the deposited mass of the radiocontaminant k in the region (g).

• The erosion rate decreases with time on out-of-core surfaces, because actinides are supposed to be continuously covered by corrosion products.

Validation of alpha simulation

The actinide simulation with OSCAR V1.4 has been adjusted to fit CATTENOM 3 PWR measurements [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF]. Fissile material dissemination occurred at cycle 8 during 231 days, with disseminating rods belonging to third cycle assemblies. The core was managed in three batches. At the end of the disseminating cycle, two batches were discharged and only the first cycle fuel batch was maintained in the core for two more cycles (cycle 9 and 10).

During nominal operation

Two calculations are presented hereafter with and without uranium dissolution in reducing conditions (during nominal operation and during end-of-cycle shutdown before primary circuit oxygenation). However, uranium dissolution in oxidizing conditions is considered for both calculations [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF].

• n° 1. "w/o U diss.": No uranium dissolution is con sidered in reducing conditions. • n° 2.

"w/ U diss.": Uranium dissolution is conside red in reducing conditions.

Calculation results are compared to the following measurements:

• 134 I primary activity, as tracer of fissile material under neutron flux, during the disseminating cycle and the seven subsequent cycles. It comes mainly from 235 U and 239 Pu fissions in the deposited fissile material. • Alpha activity measured in the primary coolant, during the disseminating cycle and the seven subsequent cycles. • Scraping on steam generator channel head surfaces, performed at the end of the disseminating cycle.

Alpha analysis of the sample provides alpha surface activity on these surfaces. However, uncertainty on this measurement is high because the scraping is realized only on a few square centimeters.

From the 134 I primary activity comparison, the following adjustments are done:

• The dissemination rate and duration are adjusted during cycle 8 to match with the increase in 134 I activity as there is no phenomenological modeling of the dissemination. Compared to calculations with the previous OSCAR version [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF], the dissemination rate has been increased by a factor 2.5. Indeed, the deposited fraction on the core surfaces is around 20% according to the corrosion product deposition rate, whereas it was assessed to 50% for actinides with the previous version. • At the end of cycle 8, two fuel batches over three are discharged. Consequently the core contamination is automatically reduced by two thirds at the beginning of cycle 9. Nevertheless an additional reducing factor is applied at beginning of cycle 9 to take into account the fact that fissile material deposition is actually not homogeneous during dissemination and occurs preferentially in the vicinity of dissemination rods. This reducing factor is determined in order to match calculated with measured 134 I activity at the beginning of cycle 9. • The specific surface for actinides in the deposit layer is adjusted to 700 cm 2 /g. This is the only adjusted parameter for deposition and erosion mechanisms applied to actinides, where four parameters were independently adjusted for actinides in the previous OSCAR version [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF]: deposition and erosion rates, on core surfaces and on out-of-core surfaces. • The erosion rate decay constant on out-of-core surfaces is adjusted to 8.10 -4 per day and 5.10 -4 per day for calculations n° 1 and n° 2, respectively.

We observe that both calculations are close to measurements for 134 I primary activity (Figure 3). It must be noted that no particular parameter adjustment is done from cycle 9 to cycle 15.

The 134 I activity decrease from one cycle to the following is due to the partial reloading of the core. After the dissemination, the 134 I activity increase along a cycle is due to:

• Evolution of the fissile material under neutron flux,

• Slow uranium dissolution (calculation n° 2),

• Transfer from the out-of-core surfaces due to erosion/deposition mechanisms. In case of uranium dissolution, the erosion rate decay constant is lower and a higher transfer from out-of-core surfaces counterbalances the uranium dissolution on the core. Indeed, dissolved uranium is trapped on the ion exchange resins and disappears from the primary circuit.

Alpha activity in the primary coolant is due to disseminated particles during cycle 8 and to eroded particles during the following cycles. Figure 4 shows that the calculated alpha activity in the primary coolant during nominal operation is consistent with the measurement. This indicates that the erosion and deposition rates applied to actinides are at the right level with very few adjustments to the corrosion product rates.

Nevertheless, alpha activity seems to become too low after a few cycles. This could be improved by increasing the size of the particle. The deposition rate would be lower, and the particle activity higher.

Lastly, the calculated alpha surface activity on steam generator's channel head is also consistent with the measured value (within 20%). It must be noted that the calculated alpha surface activity is 40% higher on steam generator tubes, due to the smaller hydraulic diameter.

All these results show the consistency between corrosion products and actinides in particulate form in nominal operation. Regarding the uranium dissolution rate, it is difficult to assess as both calculations allow good simulation of actinides. Nevertheless, a higher dissolution rate than in calculation n° 2 could not lead to corre ct 134 I activity simulation without separate adjustment of the erosion rate for core surfaces and out-of-core surfaces. Therefore the uranium dissolution rate used in calculation n° 2 is likely a maximal value. It leads to the distribution illustrated in Figure 5 between 238 Pu included in uranium oxide particles (linked form) and 238 Pu evolving separately after uranium oxide dissolution (free form). The distribution is the same for other alpha emitters. This shows that 7 cycles after dissemination, around 33% of the disseminated particles has dissolved.

According to the simulation, the alpha activity in the primary circuit is mainly in the form of uranium oxide particles, and deposited on the surfaces. The ratio of in water to deposited alpha activity is lower than 10 -4 during dissemination, and lower than 10 -6 after. Filtering campaigns carried out during end-of-cycle shutdown [START_REF] Génin | Simulation of Alpha Contamination in PWR with the OSCAR code[END_REF] show that alpha activity in the coolant increases by a factor 10 to 100 after the power shutdown and remains mainly in particulate form. Moreover, a correlation between the alpha activity and the corrosion product 60 Co in particle form is observed along the shutdown process.

These observations suggest that dissolution of the corrosion layer during shutdown induces the release of actinides into the coolant. However, the dissolution rate of corrosion products is maximal at the time of primary circuit oxygenation whereas it is not the case for alpha activity in the coolant. Therefore, the alpha release is not directly linked in the OSCAR modeling to the corrosion product dissolution rate but to an increased erosion rate during the shutdown process, accounting for deposit embrittlement due to corrosion layer partial dissolution.

In calculations n° 1 and n° 2, the erosion rate is increased during shutdown by a factor of 10 for corrosion products, and a factor of 50 for actinides. Results are presented in Figure 6 at the end of cycle 11. Activity levels for alpha emitters and 60 Co in particulate form are correctly simulated, as well as the correlation between alpha and 60 Co particulate activity along the shutdown process.

The enhanced erosion factors could have been chosen differently. Nevertheless, we can conclude that enhanced erosion during shutdown can account for increased alpha activity and for correlation between alpha emitters and corrosion products in particulate form in the primary coolant. Moreover, this enhanced erosion does not affect the radiocontaminant distribution on the different surfaces of the primary circuit: surface activity does not significantly vary during the shutdown. In case of fuel oxide dissemination, fission products are enclosed in the disseminated particles. The whole fission product spectrum may be found in the primary circuit, and not only the gaseous and volatile fission products.

From 134 Cs (2 year period) and 137 Cs (30 year period) primary activity measurements for reactors contaminated with fissile material, we can assess that fission products stay enclosed in fuel oxide particles during several cycles. Indeed, unlike 137 Cs, 134 Cs is not created from fission but from neutron capture on 133 Cs, a stable nuclide coming from fissions. In nuclear fuel, 133 Cs and 137 Cs accumulate along irradiation. Therefore the 134 Cs/ 137 Cs activity ratio in the fuel is around zero at the beginning of irradiation but gradually increases with fuel burnup. In the primary coolant, we observe that the 134 Cs/ 137 Cs measured ratio corresponds to the burnup of the disseminated fuel during several cycles. This shows that cesium is very slowly released from disseminated fuel oxide particles. If cesium was quickly released from fuel oxide particles, there would be no more 133 Cs accumulation in the primary circuit and the measured 134 Cs/ 137 Cs ratio would be very low.

Consequently, it is assumed that disseminated fission products, whatever their related element, follow the same behavior than actinides, undergoing the same erosion and deposition mechanisms. Fission products are likely released from the fuel particles at the rate of the uranium oxide dissolution. This behavior is important mainly for long life fission products, as shorter life fission products quickly disappear from fuel particles. It should be considered even for gaseous nuclides like 85 Kr (11 year period).

Fission Products emitted by recoil from the disseminated fissile material

Fission products are emitted into the primary circuit from fissions in the fissile material contamination when it is deposited under neutron flux. Once again, the whole fission product spectrum can be released. Then each fission product evolves in the primary circuit according to the behavior of the related element. This behavior is well known only for gaseous and volatile fission products, which are soluble in the primary coolant (see § 3).

However, fission products related to other elements induce negligible radioactivity in the primary circuit compared to xenon, krypton, iodine, and cesium isotopes. Moreover, if their radioactive decay period is short enough (lower than one year), they are not a concern for waste and effluent management as they disappear rapidly. On the other hand, long life fission products in the primary circuit are mostly included in fuel oxide particles rather than emitted by fissions in the actinide contamination. Therefore they can be considered as insoluble and following the same mechanisms as actinides. As an example, 99 Tc, 93 Zr, 107 Pd are treated as insoluble in the OSCAR modelling. These difficult-to-measure long life fission products are important for waste and effluent management.

Lastly, a specific adsorption/desorption model [START_REF] Génin | The OSCAR code package : a unique tool for simulating PWR contamination[END_REF] has been introduced in OSCAR to describe the strontium behavior. It is used to assess the 90 Sr (29 year period) release into the primary coolant during an end-of-cycle shutdown. This behavior is related to the pH decrease, which induces desorption of strontium trapped on the primary circuit surfaces. However, the amount of strontium involved has been directly emitted from fissions in the primary circuit, whereas strontium disseminated within fuel particles stays enclosed in these particles and does not contribute to the strontium release.

Conclusion

This paper shows that the OSCAR V1.4 code is a powerful tool for simulating fission product and actinide contamination in PWR primary circuits, and is helpful for fuel and waste management as well as for the alpha risk assessment. The main conclusions are as follows:

• The fuel code ALCYONE and its fission gas module MARGARET have been adapted to defective rod configuration and allows calculating the release in the primary circuit for gaseous and volatile radionuclides in case of a rod clad defect. Validation against a unique experimental database including defective PWR rods as well as analytical irradiations in a research reactor shows that the difference between calculated and measured fuel released fractions is lower than a factor of 2 for UO2 and MOX fuel up to 50 GWd/t. • Mechanistic erosion and deposition models have been calibrated for corrosion products and applied to actinides. For that, a specific computation scheme has been used, which allows calculating the contamination of corrosion products as well as of fission products and actinides in a PWR primary circuit with the same data set. This provides a fair simulation of actinides during nominal operation, with very few adjusted parameters. It confirms the coupled behavior of corrosion products and actinides in the primary circuit. However, the uranium dissolution rate could not be precisely assessed, as it has very little impact on alpha contamination. • The alpha activity increase in the primary coolant during end-of-cycle shutdown is correctly simulated considering enhanced erosion, as well as the correlation between alpha activity and corrosion product 60 Co in particulate form along the shutdown process. Enhanced erosion accounts for deposit embrittlement due to corrosion layer partial dissolution.

The qualification of the OSCAR code should be pursued on more reactor cases to consolidate the validation domain.

  corrosion products. This is to account for the fact that actinides are deposited at the surface of a corrosion product layer and not intimately blended in the corrosion layer. In the above expressions, spec S is the specific surface of deposited actinides (m 2 /g), w S is the wet surface of the region (m 2 ), tot m is the total deposited mass of corrosion products in the region (g) and k

Figure 3 -Figure 4 -

 34 Figure 3 -Calculation/measurement comparison for 134 I primary activity

Figure 5 -

 5 Figure 5 -Distribution between linked and free forms of 238 Pu alpha emitter calculated with OSCAR V1.4 (calculation n° 2)

Figure 6 - 4 . 4

 644 Figure 6 -Calculated and measured alpha activity in the coolant during shutdown at the end of cycle 11


Didier Demolin

Bernard Teston

published or not. The documents may come


Christian Latge
email: christian.latge@cea.fr

François Beauchamp

Leïla Gicquel

Teaching Sodium Fast Reactors in CEA

Among the Fast Neutron Reactor Systems, the SFR has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype, and commercial size reactors since the 1940s. This experience corresponds to about 402 years of operation by end of 2010. Six reactors are in operation: BOR60, BN600 and BN800 in Russia, Joyo in Japan, FBTR in India and CEFR in China. One reactor is being commissioned: PFBR (500MWe) in India and several projects are currently developed: FBR1 and 2 in India, BN1200 in Russia, JSFR in Japan, PGSFR in Korea, CFR-600 in China and ASTRID in France. In order to support operation of existing reactors, design activities for new projects and decommissioning of old reactors, it is mandatory to develop skills, more particularly among the young generation, who will operate these new reactors. In addition, education and training is also essential to share the knowledge among the teams involved in Research and Development. Several strategies are developed at the national level, or within multilateral framework, like EU or IAEA to support development of Fast Reactors. In France, to answer to this increasing demand of Education & Training, four sessions are proposed, within the frame of INSTN (French National Institute for Nuclear Science and Technology):

-SFR: History, main options, design and operational feedback -SFR: Functional analysis and design -SFR: Safety and operation -SFR: Sodium structures interactions The French Na School (ESML) provides since 1975 also several sessions dedicated to Na facilities design, safe operation, handling, and also decommissioning and sodium treatment. Beside courses, practical exercises are organized during each session. 10 different modules are available, ranging in length from 1 to 5 days. CEA contributes also to the organization of European Sessions dedicated to Sodium Fast Reactors, organized within the frame of the European Commission (CP-ESFR, ESNII+ , ESFR-SMART) This Education and Training strategy is a key element for the future of the development of Sodium Fast Reactors, and more particularly ASTRID project. CEA is ready to share training experience and to collaborate with other foreign Education and Training Entities.

I. INTRODUCTION

Among the Fast Neutron Reactor Systems, the Sodium Fast Reactor (SFR) has the most comprehensive technological basis as result of the experience gained from worldwide operation of several experimental, prototype, and commercial size reactors since the 1950s. This experience corresponds to around 420 years of operation by end of 2017. Six reactors are in operation: BOR60, BN600 and BN800 in Russia, Joyo in Japan, FBTR in India and CEFR in China. One reactor is being commissioned: PFBR (500MWe) in India and several projects are currently developed: FBR1 and 2 in India, BN1200 in Russia, JSFR in Japan, PGSFR in Korea, CFR-600 in China and ASTRID in France,, in partnership with Japan. In order to support operation of existing reactors, design activities for new projects and decommissioning of old reactors, it is mandatory to develop skills, more particularly among the young generation, who will operate these new reactors. In addition, education and training is also essential to share the knowledge among the teams involved in Research and Development. Several strategies are developed at the national level, or within multilateral framework, like EU or IAEA to support development of Fast Reactors [START_REF] Latgé | Education & Training in support to Sodium Fast Reactors around the world[END_REF] In France, the new objective is to build a GENERATION IV reactor prototype so-called ASTRID. This decision has motivated an important and rapid increase of R&D work, orientated towards the design and conceptual evaluations.

II. Education and Training at ESML

The objectives of the Sodium School (ESML) are to synthesize knowledge, to share it between CEA experimental facilities operators and consequently to support R&D activities, to train operators able to work on Sodium Fast Reactors, to train design engineers involved in SFR projects and to train fire brigades. Its role has always been to adapt its offer and its training content to the changing needs for reactor operation, experimentation and for design activities. Trainees usually belonged to French companies such as CEA, EDF, AREVA and IRSN, or any companies involved in sodium activities belonging or not to the nuclear industry. At the early stage of its creation, ESML intended to be opened to foreign countries. Specific training sessions were provided for German operators for SNR300 (1983), Japanese operators for the first start-up of Monju reactor (90's) or in support to the PFR and DFR decommissioning projects (UK). More recently, ESML in association with PHENIX plant operator has extensively increased its opening to foreign institutes, such as trainees from CIAE in China, ROSATOM in Russia on Reactor technologies, safety and operation, or IGCAR in India dedicated to Safety.

ESML provided also specific sessions to Chemical industry, such as UOP (USA).

The pedagogical approach consists in a combination of various educational means: lectures, discussions and Training on a Sodium loop. Since 1975, more than 6000 trainees have received a training at the Sodium School. The following items are currently addressed: physicochemistry of sodium coolant (physical and chemical properties), purification, corrosion, contamination, cleaning and decontamination… Sodium technology, description and operation of components, instrumentation, visualization, inspection and repair, are also presented in dedicated sessions; during these sessions, exercises involving operation and intervention procedures on the sodium loop are organized. Sodium safety is always a key part of the sessions. Specific hazards induced by the chemical properties of sodium are described: sodium-water reaction and hydrogen risk assessment, sodium fires, safety rules, prevention, intervention, exercise on a real sodium fire… In support to the processes used for decommissioning SFRs, ESML lecturers present and address the following items: some specific risks, dismantling techniques, sodium treatment, sodium waste storage, decommissioning of Na-K facilities…. Due to their high electrical conductivity, liquid metals, and in particular sodium and its alloys like Nak, allow the use of magnetohydrodynamic devices as actuators and sensors. Electro-Magnetic Pumps (EMP), Eddy Currents Flow Meter (ECFM) are one of the most widely used devices in liquid sodium circuit. In both cases, working principle is based on the interaction of an electrical current density, a magnetic flux density and the velocity field of the liquid metal. Understanding, implementing and using of such devices in liquid metal facility needs specific knowledge on applied electrotechnics and magneto-hydrodynamics (MHD). Dedicated benches will be proposed to support dispensed courses based on more theoretical notions.

Bench N°3: Na-water interaction

Hazard induced by the potential large interaction of Na and water in a Steam Generation Unit or a cleaning pit is often underlined: there is one discontinuous source of hydrogen and sodium hydroxide, inducing related hazards. Nevertheless, this reactivity with water is commonly used for the development of cleaning processes for structural material wetted with sodium, during handling operations and moreover for the conversion of the large amounts of sodium into sodium hydroxide, at the end of the reactor operation, during the decommissioning phase. A process, called NOAH, has been developed in CEA Cadarache and applied successfully to convert the primary sodium from Rapsodie, PFR, KNK-2 and Superphenix. It will be used also to process the sodium from Phenix. Thus, it has been decided to design and set-up two Sodium Water Reaction benches: a device for educative practical works (Figure 5) and another one for a demonstration of sodium treatment by hydrolysis with water. This first educational facility (geometries and volumes of the glass equipment) will be optimized with regards safety standards and designed for a pressure resistance of about 1.5 bar. abs. A dedicated bench will be devoted to the design of a spray fire device to carry out small-scale sodium fire experiments (Figure 5). It will be limited to a few grams of Na of sodium (10 g max).. Thermocouples are located inside the combustion chamber for measuring the spray temperature. High-speed camera and thermal imaging can be used to visualize the sodium fire combustion, in particular the distribution of the particles sizes in the jet during combustion, the temperature distribution, effects of sodium temperature and oxygen concentration.

INSTN is partner of ENEN (European Nuclear Energy Network). (www.enen-assoc.org )

To address more particularly SFR training needs related to the operation of SFR, it was intended to develop a SFR simulator, including the Energy Conversion System. The SIRENa simulator, developed by Reactor Studies Department, in Cadarache Research Center (France) (Figure 6) is used in engineering courses as an adjunct to lectures on safety and operation and can be used for human factor studies and human machine interface design. Training on such a simulator permits to perceive the effects of constraints governing the operation of a reactor (inertia, cons-reactions, regulations ...) or to apply theoretical knowledge (eg sub-critical approach). Within this project, the main functional specifications of the simulator and the needs of development of dynamic models, have been defined.This simulator represents a "Pool type" SFR, the most common concept selected for large SFRs. The design consists of a safety vessel, which contains the primary vessel, reactor core, intermediate heat exchanger (IHX), secondary heat exchange, control rods, balance of plant and intermediate circuit pipework. Core cooling in SFR may be achieved by forced or natural circulation. The user can modify plant parameters, to adjust various aspects of the plant configuration and to accommodate differences between specific designs. An example of one such variation could be in the number of intermediate heat exchangers that are contained within the pressure vessel, the inclusion of coolant pumps onto the pressure vessel, In pooltype SFRs, the sodium in the primary circuit does not directly exchange its heat with the coolant of the Energy Conversion System (ECS): water if the Rankine cycle is selected, gas if the Brayton cycle is selected. The main components with their associated models are included in the simulator: pumps, heat exchangers, Plant control and protection systems, Purification Systems and their cold traps, Decay Heat Removal Systems….

The simulator is capable of simulating the following standard operational events:

-Power Increase and Decrease, in this mode user can change power with ramp in range among 0% to 100% by control rod -Reactor Scram and Restart -Reactor Start-up and Heat up -Reactor Shutdown and Cooling Moreover, the user can change the speed of pumps; accordingly, simulator shows it effect on Primary heat transport, intermediate heat transport system and steam/ gas system. Simulator allows also the user to change the material of fuel, cladding and shows it affects GIF Symposium -Paris (France) -16-17 October 2018 (through thermal properties and reactivity coefficients). Simulator is able of generating malfunction like pump shutdowns, valve failures, and regulation system failures…

Figure 1 :

 1 Figure 1: Superfennec Na loop.

Figure 2 :

 2 Figure 2: MININANET facility

Figure 4

 4 Figure 4 Bench dedicated to the characterization of Na-water interaction Bench N°4: Sodium fire

Figure 5 :

 5 Figure 5: Basic principle of spray fire demonstrative facility

Figure 6 :

 6 Figure 6: SIRENa simulator for SFRs IV. Education and Training at I2EN France has an important nuclear teaching platform organized around engineering schools, universities, involving also research laboratories, technical schools and also nuclear companies or dedicated entities, for professional training. In this context, I2EN, the International Institute for Nuclear Energy set up in 2010, is federating French entities delivering high level curricula in nuclear engineering and science related to the main following items: -Nuclear safety and radiation protection -Reactor physics & nuclear engineering

  Two reactors are currently being dismantled, Phenix and Superphénix. It was therefore necessary to support these activities and promote Education and Training Initiatives. To support this requirement, ESML (Ecole du Sodium et des Métaux Liquides), EC (Ecole des Combustibles), both located in CEA-Cadarache and INSTN (Institut des Sciences et Techniques Nucléaires) are the key schools to support the development of Sodium Fast Reactors.

Nomenclature

DHRS