Marianne Girard
email: marianne.girard@cea.fr

Frédéric Dacquait

J B Génin

Gilles Ranchoux

Geoffroy Riot

Jean-Baptiste Genin

2 code simulations

The CIRENE out-of-pile loop, dedicated to the study of corrosion product transfers and deposits on core and steam generator tube surface areas, is a 27-liter mock-up of a PWR primary circuit. It is equipped with a Chemical Volume Control System which allows to settle a particle injection device and an in-line particle counting system for this test. The selected particles are zirconium (IV) oxide, iron (II, III) oxide and nickel in powder forms, and they have been chosen with different range of particle sizes from 0 to 5 µm. The Fe 3 O 4 and Ni powders are assumed to be chemically representative of the most relevant corrosion products encountered under solid form in the PWR primary fluid. The ZrO 2 powder has been chosen for its chemically inert behaviour in order to demonstrate the relevance of the injection device and of the particle counting methodology. In this experiment, the thermal-hydraulic parameters and chemical conditions are representative of a French PWR primary circuit operational conditions and series of short duration injections of the relevant fine powders have been carried out under permanent conditions with continuous in-line particle counting.

Experimental results of the CIRENE test have led to the main following conclusions:

-the first experimental results show the consistency and the reproducibility of the overall particle concentration measurements which enables to validate the particle counting methodology and the powder injection process; -the time constant representative of the particle deposition kinetics deduced from the decrease over time of particle concentrations in the CIRENE primary circuit (an exponential law of the form C = C o e -λt ) is of the same order of magnitude for fine particles as the one obtained from the CIRENE test simulations with the OSCAR V1.2 code.

The particle deposition kineticsmodelling implemented in the OSCAR V1.2 code---which is based on Beal model --proved to be validated by this CIRENE experiment by taking into account adjustments against PWR feedback.

-OBJECTIVE OF THE TEST

The understanding of mass transfers and particularly the understanding of particle transfers in PWR primary circuit is of great importance. For this purpose, this preliminary experiment undertaken in the CIRENE out-of-pile loop, has been carried out to estimate particle deposition kinetics in the primary circuit within controlled particle injections combined with in-line particle concentration measurements. Compared to experiments previously carried out on the topic with glass particles, oil droplets or lycopodium spores in air flows [START_REF] Papavergos | Review paper: particule deposition behaviour from turbulent flows[END_REF], or sand grains in water [START_REF] Beal | Correlation for the sticking probability and erosion of particles[END_REF], this test has the original feature of implementing particles of corrosion products in a PWR primary circuit. The selected particles are zirconium (IV) oxide, iron (II, III) oxide and nickel in powder forms, and they have been chosen with different range of particle sizes from 0 to 5 µm. The Fe 3 O 4 and Ni powders are assumed to be chemically representative of the most relevant corrosion products encountered under solid form in the PWR primary fluid. The ZrO 2 powder has been chosen for its chemically inert behaviour in order to demonstrate the relevance of the injection device and of the particle counting methodology. The thermal-hydraulic and chemical conditions are representative of French PWR primary circuit operational conditions. This paper presents the experimental results of this new CIRENE test in terms of the global kinetics of particle deposition and compares them with the particle deposition rate calculated by the OSCAR V1.2 contamination simulation code, implementing the Beal model [START_REF] Beal | Deposition of particles in turbulent pipe flow on channel or pipe walls[END_REF]. The OSCAR 1 code package was presented at the Québec city Québec city International Conference [START_REF] Génin | The OSCAR code package: a unique tool for simulating PWR contamination[END_REF]. This code which is developed by CEA in cooperation with EDF and AREVA-NP simulates PWR contamination by fission products, actinides and corrosion products.

-TEST OPERATION AND PARTICLE INJECTIONS IN THE CIRENE LOOP

The CIRENE loop outlines

The outlines of the CIRENE facility required for this test are presented in Figure 1 hereafter.

The CIRENE loop is a mock-up of a PWR primary system, with 4 electrically heated Zircaloy rods and a primary heat exchanger equipped with 4 steam generator tubes (SG). A thermal pressurizer and a primary circulation pump are used to maintain respectively the fluid pressure and the fluid velocity in the reactor coolant circuit (RCS). The thermal hydraulic conditions correspond to a single-phase flow regime along the core section and they are kept constant during the test. The 27-liter primary circuit is equipped with a Chemical Volume Control System (CVCS) which allows to settle both the injection and particle counting system. The powder injection device is placed at the outlet of the CVCS, where the fluid temperature is approximately 20 °C and th e pressure is 150 bar. The injection is performed based on a Dirac's pulse type injection of a predefined mass contained in a 300 cm 3 cylinder, the powder being carried by the CVCS flow via a set of valves. The particle counting system is placed at the inlet of the CVCS where the fluid temperature is approximately 20 °C and the pressure is 2 bar. The total volume of the RCS is around 26 L, the relevant wet surfaces are around 4.8 m 2 .

Test operation

The thermal hydraulic and chemical data of this test are given in Figure 2 below provides the order of magnitude of the residence time in the RCS and CVCS circuits for flow rates set respectively at 5 200 and 6 L/h: note that the residence time of a particle, from the powder injection device up to the connection onto the RCS circuit is 216 s for a 0.3 L-cylinder with a CVCS flow rate set at 6 L/h. The CVCS flow rate may range between 5 and 9 L/h.

Figure 2: Residence times of a particle (T) expected along the CIRENE loop circuits

Particle injection sequences and particle counting

Particle injections were performed as follows:

-step 1 : injections of zirconia powder with a < 5 µm-particle size in order to validate the powder injection procedure;

RCS

T T T T = 216 s

-step 2 : injections of different powders as zirconia with a ≤ 2 µm-particle size, Fe 3 O 4 with a < 1 µmparticle size and nickel with a < 3 µm-particle size.

Table 2 below provides the characteristics of the injected powders and the relevant injected masses, these masses being consistent with the expected saturation level of the particle counters, the characteristics of each injected powder (the density data) and the total RCS volume (25,7 L).

Table 2: Characteristics of the injected powders

The primary fluid was subjected to particle counting of a 10 mL volume through laser sensor counters. Two different devices were involved for these continuous in-line measurements of the primary fluid:

model #1 with a range of particle sizes from 1 to 200 µm and for a maximum concentration which may reach 1 200 000 part/10 mL ; -model #2 with a range of particle sizes from 0.5 to 20 µm and a maximum particle concentration of 120 000 part/10 mL. The in-line particle counting was carried out as close as possible to the start of the low temperature section of the CVCS so that any dissolution of particles caused by the decrease in temperature (Fe3O4 powder for example) is limited by the time factor. The CVCS line maintained at 300°C does not involve any dissolution risk fo r the particles coming from the RCS reactor primary system, however, deposition may take place on the walls due to the laminar flow (R e ~ 450).

-EXPERIMENTAL RESULTS

The in-line particle concentration measurements

A typical histogram given by the particle counter and related to a zirconia injection is presented in Figure 3 below: the histogram provides the number of particles countered in a 10 mL -primary fluid volume over time, for a specific range of particle size. These histograms based on C = f (t), provide access to the particle deposition kinetics within the RCS of the CIRENE loop according to the following approach: a) the hypothesis is, once the maximum particle concentration Co has been reached, that the only flux responsible for the change in particle concentration is the deposition sink term on the wall surfaces of the CIRENE circuits. The erosion flux, which is one of the fluxes taken into account in the global deposition profile measurement, is considered here to be negligible compared to the deposition flux in the time period following the injections; b)

we then have an equation of type C = Co.e -λ t , where λ is the time constant associated with the studied experimental deposition speed ; c)

we determine the time constant λ as follows:

-after subtracting the background measurement, the curves are plotted in relation to t o (t o corresponding to the particle concentration peak Co = Cmax) and studied over a time period during which the deposition regime is constant, i.e. globally on ∆t = 800 s for zirconia injections; -each curve is then treated as an "exponential trend curve" from which the term λ CIRENE is identified as C = Co.e -λ CIRENE t (λ is expressed as s -1

). As an example, Figure 4 for two particle size channels -Zirconia injection #4 The experimental term λ CIRENE will be compared with the time constant λ OSCAR associated with the deposition kinetics modelled in the OSCAR code and calculated in the case of the CIRENE primary circuit for a given particle size distribution. This comparative study is presented in chapter 4.

Experimental results

Preliminary test -step 1

The first injections and particles counting have led to the following points: the background level in particle concentration of the primary fluid gives a population ranging for the most part between 0.5 and 1 µm, with a total value of 24 000 part/10 mL in the [0,5_20 µm] measurement range. Thus, the measurement ranges [0.7_20µm] and [1_20µm] were selected for the following injections to correctly carry out the in-line monitoring during the powder injection sequences; the injections conducted with "chemically inert" zirconia powders made it possible to validate the powder injection procedure and to check the consistency of the in-line measurements results by particles counting; due to the consistency and the good reproducibility of the step 1 results, injection sequences of powders of various natures and with various particle sizes can be carried out.

Experimental results -step 1 and step 2

ZrO2 powder injections Table 3 below shows the time constant λ CIRENE relative to injection #7 to injection #12, for the two available zirconia powders (theoretical particle size < 5 µm and ≤ 2 µm). The study is conducted for various particle size ranges over the relevant 4 time periods after the maximum particle concentration peak. The size of the measured particles is then within the range [1_20 µm]. The following points can be noted:

λ CIRENE constant measured close to the concentration peak is in the order of 15 to 19E-04 s -1 for the finest particles (< 1.5 µm); -as the two injected zirconia powders show relatively similar particle size distributions, with an actual averaged particle size around ≤ 1.5 µm, the measurements did not show any discrepancy which may be related to a particle size difference; -for a same time period, the deposition constant shows very little variation within the studied particle size range; -the deposition constant decreases to a value close to 4 to 6E-04 s -1 in the fourth time period of the study (from 2 400 to 3 200 s), we may suppose that the contribution of the erosion flux becomes more significant in such period.

Experimental time constants deduced from similar series of zirconia injections within a [0.7_20 µm] particle measurement range have led to similar values of λ CIRENE close to the concentration peak, with 15 to 18E-04 s -1

for particles < 2 µm; as for the previous zirconia injections, the λ CIRENE constant does not appear to be sensitive to the particle size in the range of the study. The following points can be noted:

Fe 3 O 4 injections
-we obtained a certain reproducibility for the injections #21 and #22 carried out in the same conditions, with a constant λ CIRENE measured close to the concentration peak in the order of 18 to 21E-04 s -1 for the finest particles (<1.5 µm). Particles with a size > 1.5 µm follow slower deposition kinetics, with a λ CIRENE of approximately 9 to 13E-04 s -1

. We observe deposition kinetics similar to the kinetics of the zirconia powders in the 3 rd time period;

-the deposition constant given by the injection # 23 (injected mass x 4) is much higher, with a value of 34E-04 s -1 measured close to the concentration peak for the finest particles (< 1.5 µm). We observe slower deposition kinetics for larger particles, with for example λ CIRENE = 19E-04 s -1 for the particle size range from 3 to 4 µm. We again find deposition kinetics comparable to the kinetics of injections #21 and #22 from the 2 nd time period.

Ni injections

Table 5 shows the values of λ CIRENE determined from injections #25 and #26 of Ni powder. The study is carried out over 4 time periods of ~ 800 s each. The measured particle size is within the range [1_20 µm].

The following points can be noted:

-the λ CIRENE constant measured close to the concentration peak reaches 16E-04 s -1 for particles with a size < 2 µm. These particles account for ~ 92% of the total nickel powder population; -for the first 2 time periods, the constant λ CIRENE seems to increase with the particle size, a trend contrary to the one observed for the Fe 3 O 4 powder: we obtain a higher deposition constant λ CIRENE for particles of a size between 3 and 4µm, with for example, λ CIRENE = 32.5E-04 s

Summary status

The injections of the various powders led to similar deposition constant values for particle size ranging from 0.7 to 2 µm, with λ λ λ λ CIRENE = 15 to 18E-04 s -1 measured after the concentration peak. This status allows to define a reference experimental constant λ λ λ λ CIRENE reference for the following study with numerical simulations.

OSCAR V1.2 SIMULATIONS OF THE CIRENE TEST

Modelling of the CIRENE loop [5]

The schematization describes the RCS with 22 relevant control volumes and the CVCS with low temperature (T = 20°C) and low pressure (P = 2 bar) with 2 cont rol volumes simulating the powder injection sequences as follows: creation of two control volumes (0.3 L each) corresponding respectively to an iron and a nickel external injection source in particle form. Injections in the primary coolant occur at the control volume located just upstream from the core section. In this schematization, the volume of the RCS is 25.7 L and the total volume of the CVCS is 1.3 L.

Input data files

The reference input data file ---reference IDF ---integrates the thermo-hydraulic and chemical conditions of the CIRENE test with one iron injection and one nickel injection in particle forms. The zirconia injections are not simulated. The particle injections correspond exactly to the conditions of the injection #22 of Fe 3 O 4 powder and of the injection #25 of Ni powder, with the relevant injected masses of 10 and 190 mg.

The other data taken into account concern the particle size distribution data: the deposition speed given by the OSCAR calculation depends, inter alia, on the particle size distribution. Such distribution is represented according to a normal log law, with -"Stype": geometric standard deviation of the particle size distribution -"NZone": number of particle size ranges describing the normal log law distribution -"ZonePitch": pitch of the particle size ranges -"Dpart": mean geometric diameter. The specifications of the reference IDF CIRENE test are summarized in Table 6 hereafter, with "DensPart" corresponding to particle density. The calculation stops at t = 14.75 day-operation and does not include any shutdown phase. The particle size distribution profile corresponding to this reference IDF is presented as well.

Simulation results studies

The interpretation of the simulation results concerns the following data:

-the time constant of the deposition kinetics λ BEAL (in s -1

), based on BEAL model [START_REF] Beal | Deposition of particles in turbulent pipe flow on channel or pipe walls[END_REF] and calculated for the CIRENE primary circuit for a given particle size distribution. The deposition model implemented in OSCAR is the Beal model with a correction factor ---coef_Deposition Speed---applied to the deposition speed which was calibrated against the PWR feedback at 10 -1 to reproduce, at best, the particle concentration levels observed in the primary fluid in operation [START_REF] Dacquait | Simulations of corrosion product transfer with the OSCAR V1.2 code[END_REF]. The time constant of OSCAR deposition kinetics λ OSCAR is also regarded and λ OSCAR = λ BEAL x 10 -1

. These two calculated time constants can be compared with the experimental time constant λ CIRENE. over a time period following the injection.

-the fluxes of the various mechanisms involved and competing in the global deposition process, these fluxes being calculated over the complete CIRENE primary circuit. The characteristics of the OSCAR calculated decrease in iron or nickel masses in particulate form in the CIRENE primary circuit after the powder injection, are also studied.

COMPARISON BETWEEN EXPERIMENTAL RESULTS AND OSCAR V1.2 SIMULATIONS

OSCAR simulation with reference IDF

λ OSCAR via calculated "Deposition Speed" and coef_Depo Speed =10 -1 :

-related to the CIRENE primary circuit (22 control volumes), the calculated deposition constant λ OSCAR equals 18E-04 s -1

;

λ OSCAR is of the same order of magnitude as the experimental time constants λ CIRENE resulting from the first time period for smaller particles ≤ 1.5µm. This result is consistent regarding the mean geometric diameter of the particles of the reference IDF, i.e. 0.5 µm. particle masses in the RCS fluid The decrease in iron or nickel particle masses in the primary circuit given for the reference IDF is compared with a calculation done, firstly, by disabling the erosion mechanism, and secondly, by disabling the erosion and precipitation on particles mechanisms (the calculated particle concentration refers among others to these fluxes):

-Figure 5 below shows that the deposition process of the iron particles is only disturbed by the erosion flux and the precipitation on particles flux after 1 660 s after the concentration peak and after 3 200 s if we only take into account, as other fluxes, the precipitation flux on particles; -the deposition process of the nickel particles is only disturbed by the erosion and precipitation on particles fluxes after a much longer time, i.e. 4 260 s after the concentration peak. Flux studies : Figure 6 below relates to the deposition and erosion fluxes given by the calculation for the complete primary circuit around the Fe particle injection sequence: this figure shows that the particle counting methodology proposed in § 3.1 that "once the maximum particle concentration Co has been reached, the only flux responsible for the change in particle concentration is the deposition sink term on the surfaces of the reactor primary system" is verified around the concentration peak for the reference calculation. This is also true for the nickel element, for which the calculation gives a deposition/erosion fluxes ratio close to 1 000 at the particle concentration peak. All these first simulation results justify directly comparing the theoretical deposition constant λ OSCAR with the experimental constant λ CIRENE in the period following the particle concentration peak.

OSCAR simulations with various "Dpart" and "Denspart"

For the following simulations, the input data "DensPart" and "Dpart" of the reference IDF were modified to identify their impact on the λ OSCAR value given by the calculation. The values selected for "DensPart" are representative of the density of the different powders (see Table 2). The standard deviation "Stype" remains set at 1.78. Table 7 can be used to compare the results obtained for all these calculations and shows the following points:

-the mean diameter of particles "Dpart" has a strong impact on the value of λ OSCAR : for a density of 4 g/cm 3 the deposition constant λ λ λ λ OSCAR increases fivefold when "Dpart" changes from 0.5 to 2 µm;

-for particles with a mean diameter ≤ 1µm, the deposition constant λ λ λ λ OSCAR noticeably increases with the increase from 4 to 9 g/cm 3 in particle density, which is no longer the case for a population of particles with a mean diameter ≥ 1.5 µm.

Table 7: Density and Mean geometric diameter impacts on the λ OSCAR value -OSCAR V1.2 simulations

OSCAR simulations with the relevant powder characteristics

The density "DensPart" and the particle size distribution via the parameters "Dpart" and "Stype" of the relevant nickel and Fe 3 O 4 powders injected were taken into account for the next calculations. Both parameters, "Dpart" and "Stype", were deduced from the study of the particle size histograms obtained by the in-line counting during injection # 25 of nickel powder and injections #21 and #23 of Fe 3 O 4 powder. The time constants λ OSCAR obtained according to these new IDF are presented in Table 8. The particle distribution profiles thus defined are plotted and compared to the reference profile as well. Table 8: OSCAR V1.2 simulations with injected powder-IDF and relevant particle size distributions

The following points may be noted:

λ Ni > λ CIRENE reference : the time constant calculated for the nickel powder is the highest, with 32E-04 s -1

, which corresponds among others to the increasing in the particle mean diameter and in the particle density in the IDF. The experimental measurement gives a similar constant for a range of larger particles, between 3 and 4 µm (see Table 5); -λ Fe3O4 < λ CIRENE reference : the time constant calculated for the Fe 3 O 4 powder, i.e. 10E-04 s -1 , is similar to the experimental constant obtained with larger particles, at least equal to 1.5 µm (see Table 4);

OSCAR simulations with various ranges of particle sizes

The following study concerns the variation of the λ OSCAR time constant with specific particle size ranges which are defined reducing "Stype" and considering different "Dpart" of the particle size distributions. Figure 7 below provides the various size distributions thus defined: size range [1_1.5 µm] for the study of the Fe 3 O 4 powder, [1_2 µm] for the study of the Ni powder, [2_3 µm] and [3_4 µm] for all the powders. The experimental study undertaken for a 0-800 s time-period covers these same size ranges. The λ CIRENE /λ OSCAR comparisons are reported in Table 9 hereafter.

Figure 7: particle size distributions of OSCAR IDF for the particle size range study Concerning the Ni powder, λ CIRENE follows the increase observed for λ OSCAR but to a much lesser extent since λ CIRENE increases at the most by a factor 1.4 while λ OSCAR increases by a factor 13 within the enhance of the particle size range. Out of the three Fe 3 O 4 powder injections studied, λ CIRENE does not follow the evolution given by the calculation:

λ CIRENE remains quite stable and even decreases with the increase in the particle size range, while λ OSCAR increases by a factor 22.

CONCLUSIONS

A new experiment concerning particle deposition kinetics in a PWR primary circuit has been undertaken in the CIRENE out-of-pile loop. This test involved controlled particle injections combined with in-line particle concentration measurements. This paper presents the experimental results in terms of global particle deposition kinetics and compares them with the particle deposition rate calculated by the OSCAR V1.2 contamination simulation code, implementing the Beal model and an adjustment against PWR feedback. The CIRENE out-of pile loop is a 27-liter mock-up of a PWR primary circuit. It is equipped with a Chemical Volume Control System through which a particle injection device and an in-line particle counting system have been implemented for this test. The injected particles are zirconium (IV) oxide, iron (II, III) oxide and nickel in powder forms, and they have been chosen with different range of particle sizes from 0 to 5 µm.

Deposition time constant measurements

ZrO 2 powder injections: these series of injections lead to reproducible results giving a λ CIRENE constant close to the concentration peak in the order of 15 to 18E-04 s -1 for particle sizes ranging from 0.7 and 2 µm. For a same time period, the deposition constants do not change with the particle size ranges. Fe 3 O 4 and Ni powder injections:

-a series of Fe 3 O 4 powder injections leads to reproducible results giving a constant λ CIRENE close to the concentration peak in the order of 18 to 21E-04 s -1 for the finest particles with a size ranging from 1 and 1.5 µm;

-the Ni powder injections give a relevant λ CIRENE value in the order of 16E-04 s -1 for particles size ranging from 1 and 2 µm; -some dependency of the λ CIRENE deposition constant on the size of particles is observed in the 1 st time period, but with opposite trends for the two powders : thus, we observe an upward trend in λ CIRENE_Ni and on the contrary a downward trend in λ CIRENE Fe3O4 when the particle size increases.

Comparison between experimental results and OSCAR V1.2 simulations

Simulations with the reference-Input data file (IDF): a good agreement between the OSCAR calculated deposition kinetics and the experimental kinetics λ CIRENE reference is obtained considering a population of fine particles (0.7-1.5 µm) in a 0-800 s time period following the concentration peak. Furthermore this first simulation points out that, as for the PWR simulations, the kinetics constant λ BEAL (in s -1

) deduced from the calculated "Deposition Speed" values must be lowered applying a 10 -1 correction factor.

Simulations with the relevant powder characteristics-IDF: the deposition constants given by the calculation remain on the whole in the same order of magnitude as the constants given by the experiment. However, we note a better agreement between calculations and experiment results for the larger experimental particle size ranges;

Simulations with various ranges of particle sizes-IDF: OSCAR simulations point out that λ OSCAR increases from a factor 13 for Ni powder, to a factor 22 for Fe 3 O 4 powder while the particle size range enhances. No experimental dependency of the deposition kinetics with the particle size range emerges as strong as in the calculation. The best correspondence between calculations and experiment results is obtained for the [1-1.5 µm] range for the nickel injection and for the [1-2 µm] range for the iron injection;

The calculated fluxes study indicates that the deposition process of the injected particles is only disturbed by the erosion and precipitation fluxes on particles after 1 660 s after the concentration peak for iron particles, and after a relevant much longer time of 4 260 s for the nickel particles. All these results justify studying the experimental time constant λ CIRENE in the time period following the particle concentration peak. The particle deposition kinetics modelling implemented in the OSCAR V1.2 code---which is based on Beal model ---proved to be validated by this CIRENE experiment by taking into account adjustments against PWR feedback.

OSCAR 1 :Figure 1 :

 11 Figure 1: the CIRENE loop outlines required for powder injection and particle counting

Figure 3 :

 3 Figure 3: Typical histogram C = f (t) provided by the particle counter settled on the CIRENE CVCS -Zirconia injection #7

Figure 4 :

 4 Figure 4: Determination of the time constant λ CIRENE through exponential trend curves C = Co.e -λ t

Figure 5 :

 5 Figure 5: Fe mass in the RCS fluid calculated around the Fe particle injection with different disabled mechanisms (erosion mechanism and precipitation on particles mechanism)

Figure 6 :

 6 Figure 6: erosion and deposition Fe fluxes intensities calculated around the injection at 12.75 days

  deposition and erosion fluxes in Fe in the RCS around the Fe_ part injection OSCAR V1.2 -reference IDF

Table

Table 1 :

 1 . The thermal hydraulic parameters which remain stable until the shutdown, lead to a R e number (Reynolds number) of ~ 2E+05 in the core and the SG tube sections with a relevant 5 200 L/h primary flow rate. The chemical conditioning of the primary water is representative of PWR operation. Thermal hydraulic parameters and chemical conditioning of the CIRENE test

	Thermal hydraulic parameters
	Pressure		15 MPa (150 bar)
	Primary flow rate	1.46E-03 m	3 /s (5 200 L/h)
	Mass flow rate in the core section		2 640 kg/m	2 /s
	Core section -inlet / outlet temperature		297°C / 320°C
	-Re		1.8E+05
	-fluid velocity		3.6 m/s
	Primary fluid chemical conditioning	B = 100 mg/kg Li = 0.6 mg/kg pH 300°C = 7.35
		H 2 = 12-28 cm	3 /kg O 2 < 10 µg/kg

Table 3 :

 3 Experimental time constants -λ CIRENE -deduced from zirconia powder injections

	injection #		inj_7	inj_8	inj_9	inj_12
	injected powder : ZrO 2		ZrO2 < 5 µm ZrO2 < 5 µm ZrO2 ≤ 2 µm ZrO2 ≤ 2 µm
	λ (s-1)	part >1µm	14-15E-04	17E-04	18E-04	16E-04
	1rst period (0-800 s)	1<part<1.5µm	15E-04	17E-04	19E-04	16E-04
		3<part<4µm	12E-04	17E-04	16E-04	15E-04
	λ (s-1)	part >1µm	11E-04	11E-04	10E-04	9E-04
	2d period (800-1 600 s)	1µm<part<1.5µm	11E-04	11E-04	10E-04	9E-04
		3µm<part<4µm	8E-04	11E-04	10E-04	8E-04
	λ (s-1)	part >1µm	7E-04	7E-04	7E-04	6E-04
	3rd period (1 600-2 400s) 1µm<part<1.5µm	7E-04	7E-04	7E-04	6E-04
		3µm<part<4µm	6E-04	7E-04	7E-04	7E-04
	λ (s-1)	part >1µm	4E-04	5E-04	6E-04	5E-04
	4th period (2 400-3 200s) 1µm<part<1.5µm	5E-04	6E-04	6E-04	4E-04
		3µm<part<4µm		6E-04	7E-04	5E-04
	in-line particle counter #1 [1_200 µm] Cmax. = 1 200 000 part/10ml

Table 4

 4

	injection # (injected mass)	inj_21 (m)	inj_22 (m)	inj_23 (m x 4)
	injected powder : Fe 3 O 4 <1µm
	λ (s-1)	part>1µm	18E-04	18E-04	33-34E-04
	1 rst period (0-800 s)	1< part < 1.5µm	19E-04	21E-04	34E-04
		1,5µm < part < 2µm	11E-04	13E-04	27E-04
		2 < part < 3µm	10E-04	11E-04	21E-04
		3 < part < 4µm	9E-04	12E-04	19E-04
	λ (s-1)	part>1µm	7E-04	8E-04	8-9E-04
	2 d period (800-1 600 s)	1µm< part < 2µm	7E-04		10E-04
		2µm< part < 3µm	7E-04		9E-04
		3µm< part < 4µm	8E-04		9E-04
	λ (s-1)	part>1µm	6E-04	2E-04	6E-04
	3 rd period (1 600-2 400 s) 1µm< part < 2µm	6-7E-04		5-6E-04
		2µm< part < 3µm	6E-04		7E-04
		3µm< part < 4µm	7E-04		5E-04
	λ (s-1)	part>1µm	2E-04	3E-04	3E-04
	4 th period (2 400-4 200s) 1µm< part < 2µm	2E-04
		2µm< part < 3µm	1E-04
		3µm< part < 4µm	1E-04

shows the λ CIRENE values determined from the injections #21 to #23. The study is carried out over 4 time periods. The size of particles measured is within the range [1_20 µm]. in-line particle counter #1 [1_200 µm] Cmax. = 1 200 000 part/10ml

Table 4 :

 4 Experimental time constant λ CIRENE deduced from Fe 3 O 4 powder injections

Table 5 :

 5 Experimental time constant λ CIRENE deduced from Ni powder injections

	-1 in the 2	nd time period

Table 6 :

 6 Reference OSCAR IDF of the CIRENE test with the relevant particle size distribution data

deposition flux_reference IDF erosion flux_reference IDF Fe_part injection Evolution of the Fe_particle mass in the RCS primary fluid OSCAR V1.2_reference IDF

			mFepart (mg)
			1,E+01
								1 660 s
			1,E-01
										3 200 s
												reference
			1,E-03								without erosion mechanism
												without erosion and precipitation
				853 s							on particles mechanisms
			1,E-05										t (day)
				12,73	12,74	12,75		12,76	12,77	12,78	12,79	12,8	12,81	12,82	12,83
			Particle size distributions with various Dpart and Stype
	particle size 2 000 000 distribution
							reference IDF Dpart : 0.
	1 600 000			flux (g/s)
				1,E-04
	1 200 000
	800 000
				1,E-05
	400 000
										particle diameter
												(µm)
	0
	0 , 0	0 , 5	1 , 0	1,E-06 1 , 5	2 , 0	2 , 5	3 , 0	3 , 5	4 , 0	4 , 5	5 , 0	5 , 5	6 , 0
														t (in day)
				1,E-07
				12,70		12,72	12,74	12,76	12,78	12,80	12,82	12,84

5µm Stype : 1.78 Fe3O4 powder Dpart : 1µm Stype : 1.4 Ni powder Dpart : 1.2µm Stype : 1.4

			calculated deposition time constant λ OSCAR (s -1 )
	Particle characteristics	Mean geometric diameter "Dpart"
	reference IDF		0.5 µm	1 µm		1.5 µm	2 µm
	Density	4 g/cm3	18E-04	56E-04		83E-04	100E-04
	"DensPart"	6 g/cm3	22E-04	62E-04		85E-04
		9 g/cm3	27E-04	65E-04		79E-04	74E-04
						Particle size distribution	OSCAR
				"Dpart"	"DensPart"	Stype	(s	-1 )
			reference IDF	0.5 µm	4 g/cm3	1.78	18E-04
			Fe3O4 powder IDF	1 µm	4 g/cm3	1.4	10E-04
			Ni powder IDF	1.2 µm	9 g/cm3	1.4	32E-04

Table 9 :

 9 calculated λ OSCAR and experimental λ CIRENE time constants according to particle size ranges

	Ni powder λ OSCAR (s-1)	0.5 µm* 18E-04	1.4µm 9E-04	2.5µm 94E-04	3.5µm 120E-04	Fe 3 O 4 powder λ OSCAR (s-1)	0.5 µm* 18E-04	1.2µm 5E-04	2.5µm 37E-04	3.5µm 110E-04
	0 -800 s		particle size range experimental measurements	0 -800 s	particle size range experimental measurements
			> 1µm	1_2µm	2_3µm	3_4µm		> 1µm	1_1.5µm	2_3µm	3_4µm
	λ CIRENE inj_26 17E-04	17E-04	20E-04	23E-04	λ CIRENE inj_21	18E-04	19E-04	10E-04	9E-04
	(s-1)	inj_25	16E-04	16E-04	16E-04	18E-04	(s-1) inj_22	20E-04	21E-04	11E-04	12E-04
	*reference IDF : Stype = 1.78				inj_23	34E-04	34E-04	21E-04	19E-04

"Dpart" OSCAR simulations (Stype : 1.1) "Dpart" OSCAR simulations (Stype : 1.1)

RCS

Core section 4 Zry4 claddings 320 °C/150 bar 297 °C/150 bar Secondary circuit 690TT tubes 297 °C/150 bar 20 °C/2 bar CVCS CVCS pump 5 -9L/h


Gilles Ranchoux
email: gilles.ranchoux@edf.fr

Matthieu Wintergerst

Martin Bachet

Stéphanie Leclercq

OPTIMIZED HIGH TEMPERATURE OXIDATION AND CLEANING AT BUGEY 3

As a part of the EDF Source Term Reduction project, an experimental procedure was carried out at BUGEY 3 further to the steam generator replacement. This innovative procedure consists in theory in two complementary phases /1/:

 Phase 1: a SG tubes optimized oxidation performed during pre-critical hot functional tests (basic and reducing chemistry) aims to generate an as protective as possible inner oxide layer allowing to reduce the later nickel release,  Phase 2: a cleaning procedure of the primary circuit performed under acid and reducing chemical conditioning at 170°C intends to dissolve and eliminate the outer oxide layer by a simultaneous purification.

The objective of such a procedure is to reduce corrosion products inventory (mainly nickel) generated by the first SG tube oxidation during hot functional tests and first operation months by carrying out an appropriate cleaning procedure.

Gains were expected not only on RCS and auxiliary systems contamination, dose rates and thus collective dose but also on next outages duration.

The objective of this paper is to describe the process implementation at BUGEY 3: effective procedure put in place, monitoring program (chemistry and dose rate measurements, EMECC campaign) and firsts results.

INTRODUCTION

The reduction of collective dose is an important objective for all nuclear operators. This consists not only in improving the reactor shutdown organization (time spent in control area, biological shielding, …) but also in improving the radiological state of the unit and the efficiency of the source term reduction operations.

Significant efforts are continuously performed to contribute to source term and contamination reduction in primary and auxiliary circuits. Actions put in place at EDF concerns at the same time preventive ways (operational procedures optimization, purification media optimization) /1/ and curative ones (decontamination).

One of the most investigated way to reduce source terme preventively is the first Steam Generator (SG) tubes oxidation for a new built unit or after Steam Generator Replacement (SGR) for units already in operation. As a part of the EDF Source Term Reduction project, an innovative procedure was proposed by EDF/R&D department in 2010 thanks to the accumulated knowledge on metal release and dissolution mechanism of Ni-based alloy SG tubes /2/. This procedure was carried out at Bugey 3 further to the SGR. Due to an induced low nickel inventory and thus to a low 58 Co activity, gains were expected not only on next outages duration (shorter purification time before the last RCP shutdown) but also mainly on RCS and auxiliary systems contamination, dose rates and thus collective dose.

The objective of this paper is to describe the process implementation at BUGEY 3 : effective procedure put in place, monitoring program (chemistry and dose rate measurements, EMECC campaign) and first results.

PROCEDURE DESCRIPTION

Theoretical reminder /2/

When they are first exposed to the primary coolant at high temperatures, nickel-based alloys get covered by an inner oxide layer (rich in chromium) that strongly slows down their ulterior generalized corrosion and an outer oxide layer providing a significant amount of nickel available for dissolution. For Thermal Treated 690 alloy equipping recently built steam generator, this phenomenon can become extremly low after several cycles, which makes the nickel source term in the primary circuit really small and by consequence little 58 Co is generated by activation of natural nickel.

Many studies aimed at determining the best chemistry to get a very protective passive layer and it has long been recognized that an important mass of nickel is oxidized during its formation. Oxidized nickel is not immediately released to water because at high temperture, nickel oxide, hydroxide and ferrite have very low solubility and precipitate, giving the well known chromium-poor external oxide. However, the release of nickel is not inhibited, but only delayed. It is therefore desirable to clean nickel out of the primary circuit, but despite a few trials, especially during Sizewell B commissionning, this cleaning has probably never been really efficient.

The new idea developed by EDF R&D department is that the cleaning of the primary circuit (mainly SG surfaces) can be achieved by an acid and reducing chemistry at a temperature close to 170°C, which would enable the dissolution of nickel containing oxides, while preserving the chromium rich passive layer on primary circuit surfaces. This temperature has been chosen as the result of a compromise between thermodynamic (nickel solubility decreases when temperature increases) and kinetics (dissolution kinetic increases with temperature). Before this cleaning step, a pre-oxidation of the tubes is necessary, and should lead to a stable and as protective as possible passive layer, while minimizing the mass of nickel made available for later release.

The main benefit associated with this cleaning would be to avoid elevated 58 Co oxygenation peaks activities following new steam generators operation, and to obtain quickly a plant with both low release rates and low nickel inventory. Other effects will probably be a reduction in dose rates, no more need for prolonged purification steps during shutdown and a generally cleaner reactor.

The optimum conditions for dissolving nickel after the pre-oxidation step should focus on the different possible nickel-bearing phases, which are predicted to be slightly different just after commisioning new steam generator to what is usually encountered during PWR shutdowns. Based on the general knowledge on the behavior of corrosion products, a compromise should be made between solubility, which is higher at lower temperature and dissolution kinetics which show the opposite trend. Acidic pH is favorable for the dissolution of all phases, and hydrogen is required for the dissolution of nickel ferrites. Hydrogen also provide protection against the dissolution of the chromium oxide layer responsible for the slow corrosion rate. The consequences of such a shift were estimated to be negligible provided that the second phase ended about 8 month before the following outage. As a matter of fact, this time is enough to allow a significant radioactive decay of the 58 Co (half life : 70.8 d) generated between the first oxidation and the cleaning phase. Moreover, it is likely that the supplementary oxidation time during operation (more than 7 months) leads to a more stable protective inner oxide layer than for the theoretical procedure.

Furthermore, a high purification rate is required during the cleaning phase in order to assure a fast nickel elimination. For safety reasons (mainly RHRS thermal fatigue), it has not been possible to stay in a RHRS cooling state for 200 hours taking advantage of the same purification rate as during oxygenation (27 t/h). A steam generator cooling state has been prefered but imposed to stay with a primary pressure about 80 bars. At that pressure, the minimal temperature acceptable for safety reasons is about 190°C, temperature which is slighty higher than the ideal temperature of 170°C. However, considering available nickel solubility data, it has been estimated that the efficiency loss caused by a 20°C deviation from the ideal temperature would be offset by a high purification rate ( 20 t/h) and a higher dissolution kinetic.

The schematic diagaram of the effective procedure is shown on figure 2. Moreover, a decisional flow chart has been elaborated in order to stop the procedure if efficiency criteria (mainly nickel concentration) were not reached. The minimal nickel concentration above which a significant efficiency of the procedure could be expected has been set to 1 ppm.

Expected gains on outage collective dose have been estimated to about 30% from hypothesis based on the more updated international theoretical and experimental feedback concerning on one hand nickel solubility and dissolution kinetic and on the other hand oxide layer structure caracterization :

 20% are attributed to nickel oxide and/or nickel hydroxide elimination "directly" at the root on steam generator tubes ;

 10% are attributed to nickel ferrite generated on the tube walls on steam generator but also on the other parts of the primary circuit (hot, cold and crossover legs).

A lower 58 Co oxygenation peak activity is also expected due to a significant decrease of the nickel mass potentially activable under neutronic flux. However, availibility gains due to a shorter purification time between oxygenation and the reach of the last RCP shutdown criteria ( 58 Co activity = 50 GBq/t) are low with regards to an incompressible time below which it is not possible to avoid some obligatory simultaneous works during the outage. This procedure, as proposed above, had never been experimented as by EDF or by any other operators. However, although innovative, this does not use new chemical conditionning likely to cause damages on the primary circuit (stress corrosion cracking, fuel damage, axial offset anomaly). In fact, chemical specifications are totally respected (boron, lithium and hydrogen concentration) for both phases. Moreover, this procedure is composed by several steps which have all been rolled out independently by EDF or at an international level.

RESULTS OF THE MONITORING PROGRAM

In addition to the current monitoring program, specific supplementary measurements (nickel and iron concentration, dose rates and gamma spectrometry) were carried out especially during the second acid and reducing cleaning phase. This section presents the different results and gives a preliminary analysis. As shown in that graph, the nickel concentration did not reach the criterion value of 1 ppm above which a significant efficiency would have been expected. Thus, it has been decided to stop the cleaning phase after only 36 hours to initial 200 hours) with respect to the operational flow chart. During the acid and reducing cleaning, one can observe that the nickel concentration increased up to 190 ppb consequently to the lithium concentration decrease and stabilised at about 130 ppb after boron dilution (necessary to reach the hot shutdown boron concentration in preparation of a quick restart). Approximately 150 g of nickel have been removed from the primary circuit which is not much compared to the expected 5 kg.

With regard to these results, two phenomena have to be explained: the nickel concentration variation and the low nickel concentration level.

Considering nickel variation, the concentration follows calculated solubility trends as a function of pH. One can observe:

 an increase when the lithium concentration decreases at constant boron concentration,  a quick decrease when the boron concentration decreases,  a stabilization when the lithium, boron and hydrogen concentration are stable.

Considering the nickel concentration level, the concentration measured during the cleaning phase stays lower than the value expected by EDF. Two complementary studies are in progress to better understand this unexpected behaviour: the chemical species present in the primary circuit on one hand, and the effective cleaning temperature upper than the temperature for which a feedback was available on the other hand.

The procedure has been performed at a temperature between 190°C and 195°C instead of the ideal temperature of 170°C for optimizing the purification rate. Solubilities are known to decrease when the temperature grows from 170°C to 190°C. Nevertheless, this aspect had been taken into account during the procedure design studies and such a solubility decrease remained acceptable as regards the expected gain on the purification rate, dissolution kinetic and the fact that the cleaning procedure acted on the nickel contained into oxides NiFe 2 O 4 , NiO and Ni(OH) 2 .

With regards to the information from chemical analysis, one can point out that the species present in the primary coolant seems to be mainly under a nickel ferrite form (NiFe 2 O 4 ) instead of a more soluble nickel oxide (NiO). In fact, iron concentration is approximately twice as high as nickel concentration which is consistent with the nickel ferrite stoichiometric ratio. A better balanced ratio between both chemical species was expected considering the usual nickel excess compared to iron when steam generator tubes are made of 690 alloy.

Figure 5: Ni and Fe concentration during the cleaning phase

In conclusion, a temperature between 190°C and 195°C seems to be not adapted to an efficient cleaning procedure under acid and reducing conditioning, although this is probably not the only reason of the observed low dissolution rate. Furthermore, specific nickel measurements have been made during the following outage RHRS connection phase (at about 170°C) and will allow to determine if the procedure would be more efficient at the ideal temperature. The origin of that phenomenon is not explained at the moment but is still under review by EDF and its partners.

Deposited activities (EMECC campaign)

In addition to chemical and volume activity measurements, EMECC gamma spectrometry measurements have been performed to determine the impact of the procedure on the surface activity on wall pipes and on steam generator tubes.

EMECC campaigns (performed by CEA for EDF needs) have been commissioned for more than 30 years on French fleet units in order to better characterize contamination mechanisms. At the same time, EDF has also commissioned and financed EMECC campaigns on foreign units (Doel, Sizewell, Trillo during the 4 last years /4/) with the contribution of several European operators in order to compare different good international practices design impact.

That device /3/ allows non destructive surface activity by gamma spectrometry measurements (HP germanium detector cooled by liquid nitrogen) based on a photon flux measurement converted in activity with the help of a pre-computed transfer function (MERCURE code /5/). A very large range of geometries can be measured as primary circuit (cold and hot side steam generator tube bundle, hot, cold and cross-over legs), CVCS (letdown line, regenerative heat exchangers), RHRS (heat exchanger, piping), … The volume activity contribution can be removed in order to focus only on the deposited activity.

CONCLUSION

After the steam generator replacement in 2010 at Bugey 3 NPP, an experimental SG tubes optimized first oxidation procedure during hot functional tests followed by a cleaning RCS phase have been rolled out. This ideal procedure consists in two successive phases:

 the "optimized first oxidation" phase consists in an optimization of the chemical conditioning during the reactor start-up under basic and reducing conditioning at hot temperature,  the "cleaning" phase consists in a acid and reducing conditioning at about 170°C during a dedicated outage.

With regards to operational aspects (safety and reactor control), the procedure had to be realised in two time separated phases.

In theory, such a procedure aims to reduce the inventory in corrosion products (specifically nickel) with expected gains on circuit contamination of about 30%.

A specific monitoring program (metal concentration measurements, volume and deposited activity by gamma spectrometry and dose rates) has been performed to estimate in real time the procedure efficiency. Thus, the nickel concentration variation, lower than the defined criterion of 1 ppm, led to stop the procedure after only 36 hours instead of the initially planned 200 hours. The contamination characterization by gamma spectrometry (EMECC campaign) and the dose rate monitoring have not allowed identifying a positive impact of the procedure on surface contamination.

The numerous data obtained during the experimentation have to be investigated more in depth in order to explain the observed phenomena and to understand why this chemical treatment has not led to the expected gains as much as for nickel elimination as for contamination and dose rate reduction. Research and development actions are already in progress:

 firstly, to better understand the impact of this procedure in terms of nickel release rate (BOREAL loop experiments /5/),  secondly, to reduce the uncertainties on nickel, nickel oxide and nickel ferrite solubilities at medium or hot temperature,  finally, to acquire new data about dissolution kinetics of nickel, nickel oxide and nickel ferrite.

The final objective is to decide if such a procedure can be viable in operation if some adjustments are proposed to improve the treatment efficiency with a view to an eventual deployment for future SG replacements of the EDF fleet and new-built nuclear power plants start-up. Beyond these technical questions not solved yet, this experimentation highlights the EDF capacity to bring on innovative procedures via its R&D department and to implement them to an industrial and operational level. Finally, even if the results have not been as good as expected, the wide mobilization of EDF teams and of its partners on that subject demonstrates the EDF will to improve radiological state of its facilities and as a consequence, the working conditions of its staff and its external contractors.

Figure 1 :

 1 Figure 1 : Schematic representation of the oxide grown in high temperature primary water on tubes made of alloy 690TT /2/

Figure 2 :

 2 Figure 2 : Description of the theoretical procedureSeveral operational issues led to adapt the theoretical procedure to the plant schedule. First of all, the restart of the plant after SGR (thus, optimized oxidation during hot functional tests) was planned in december 2010. For questions of electricity supply during winter, it has not been possible to perform the second phase immedialty after the first one. The first optimized oxidation has been limited to the hot functional tests length (136 h) and continued during operation. The cleaning phase had to be shifted 7.5 months later during summer 2011.

Figure 3 :

 3 Figure 3 : Description of the effective operational procedure

Figure 4

 4 Figure 4 shows the variation of the main operational parameters during this specific outage (boron, lithium and hydrogen concentration, temperature, purification rate, 58 Co volume activity and nickel concentration) for the cleaning phase.

Figure 4 :

 4 Figure 4 : Operational parameters during the cleaning procedure outage

Figure 7

 7 Figure 7 and 8 shows the evolution of the 58 Co and 60 Co surface activities deposited on hot legs, crossover legs and steam generator tubes (cold side) and normalized to 1 at the beginning of the cleaning phase. One can observe a slight decrease trend for 58 Co surface activity but not significant because lower than the measurement uncertainty. Moreover the same trend is not observed for 60 Co surface activity.Figure 6: EMECC device

Figure 6 :

 6 EMECC deviceThese results confirm that this specific cleaning procedure did not allow removing contamination from out-of-flux surfaces (wall pipes and steam generator tubes surface) and are consistent with the low removal of activated corrosion products inferred from coolant sampling.

Figure 7 :

 7 Figure 7: Evolution of the 58 Co surface activity on legs and SG tubes during the cleaning phase

Figure 8 :

 8 Figure 8: Normalized dose rate measured during the cleaning procedure

Co activity has decreased naturally by 20% and by a probably low contamination level on legs before SGR (no EMECC campaign before SGR). In other words, apart from any chemical consideration, this very long outage led to the complete reset of the 58 Co deposited activity and to a significant decrease of 60 Co deposited activity. A 20 month long storage of the fuel assemblies before refueling may have also led to a low dissolution of the oxide rod deposit.

Dose rates measurements

In order to monitor the eventual impact of the cleaning phase on contamination levels, two shielded dose rate measurement sets have also been performed in parallel all along the duration procedure on hot legs, crossover legs and steam generator tubes. As for surface activities, even if a slight decrease trend could be assumed in the


Thomas Jobert
email: thomas.jobert@edf.fr

Sophie Raphaël

Denis Pombet

Frédéric Dacquait

' Ascenzo

Ludovic Vaillant

Inès Dignito

Thomas Chirent

Lucie D ' Ascenzo

New status of the primary system contamination of the French PWRs -An updated analysis of the EMECC data

Keywords: surface contamination, EMECC, Co-58, Co-60, PWR, EDF, zinc, manufacturing process, steam generator replacement, OSCAR

à la diffusion de documents scientifiques de niveau recherche, publiés ou non, émanant des établissements d'enseignement et de recherche français ou étrangers, des laboratoires publics ou privés.

Introduction

More than 15 years after the presentation of the status of the primary system contamination of French PWRs, it was high time to update this presentation taking into account the new data collected through EMECC measurement campaigns on the French fleet and to describe the "new status of the reactor cooling system (RCS) contamination of the French fleet".

EMECC measurement campaigns

The EMECC campaigns consist in measuring the activities deposited inside some of the components (pipes, steam generator tubing, heat exchangers...) of the primary and auxiliary systems. The surface activities are measured by gamma spectrometry measurements using the so-called EMECC device designed by the CEA ( [START_REF] Eimecke | Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC)[END_REF]) and equipped with a High Purity Germanium detector. The background noise and the volume activities are taken into account for each EMECC measurement in order to determine the surface activities only ( [START_REF] Dacquait | Ag-110m behaviour in PWRs: Lessons learnt from the EMECC campaigns[END_REF]). Surface activities inside primary circuits of the French fleet are thus collected since 1971.

As explained in [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], complementarily to dose rate measurements performed at each outage (around primary and auxiliary circuits), EDF selects units, in the frame of a plurennial program, on which EMECC measurement campaigns are conducted. Currently, about 8 EMECC measurement campaigns are carried out every year amongst the 3 reactor series of the French fleet. Through EMECC measurements, EDF is able to gather accurate data on the surface contamination of each reactor of the fleet and can improve its understanding of contamination mechanisms. The surface activities spectrum obtained with the EMECC device is typically composed of the following radionuclides: Co-58, Co-60, Cr-51, Ag-110m, Sb-124, Fe-59, Mn-54, Zr-95, Nb-95, Zn-65. In addition to these activated corrosion products, fission products, such as Cs-134, Cs-137, I-131, Te-132, are sometimes detected.

In order to get a fairly comprehensive cartography of the surface contamination of the circuits, measurements points are located:

-On the primary circuit: primary coolant pipes (hot, crossover and cold legs) and Steam Generator (SG) tube bundle of the primary system loops, -On the auxiliary circuits: heat exchangers and pipes of the Chemical and Volume Control System (CVCS), the Residual Heat Removal System (RHRS), the Nuclear Sampling System (NSS)…

The main objectives of the EMECC measurements are:

-To assess the impact of operating practices or parameters on the surface contamination, and indirectly on the radiation field (impact of dissolved H2 content, zinc injection, B/Li coordina-tion…). -To understand contamination mechanisms and elaborate strategies to cope with said mechanisms, either by preventive or curative means. This includes shutdown strategies, as EMECC measurements are performed before and after oxygenation, -To identify the origins of pollution and develop means (material substitutions, design changes, …) to limit the productions of detrimental radionuclides, -To develop, calibrate and validate calculation tools to model and predict the contamination phenomena: the CEA has developed, in the frame of a tripartite Institute (CEA-EDF-Framatome), the OSCAR code, which latest version is the V1.4a ( [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]). -To gather a sufficient amount of data in order to launch data science-based analyses in the near future.

French PWR contamination

In this paper, we will focus on radiocobalts Co-58 and Co-60, which are the main contributors to the dose rates around the PWR primary system and the dosimetry on the French fleet. As explained in the introduction, one of the objectives of the work presented here is to compare the current trend of the RCS contamination to the previous results displayed in [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], which also focused on these 2 radionuclides.

The data taken into account here correspond to 226 campaigns on 900 MWe PWRs (CP0 and CP1-CP2 sub-series), 82 campaigns on 1300 MWe PWRs (PQY) and 23 campaigns on 1450 MWe PWRs (N4).

As in paper [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], the figures presenting the contamination status of the French PWRs (Figure 1 to Figure 3 and Figure 6 to Figure 8) show the ranges of Co-58 and Co-60 deposited activities in the lot legs (HL), cold legs (CL), cross-over legs (COL) and SG tubing of the RCS given the cycles of the reactors for the different standardized plant series.

For each type of PWR, the measured deposited activities are between the "Min" and the "Max" curves. The average of the surface contamination activities is also pointed out, for each cycle of each series.

In the case of the 900 MWe series, the data are separated into 2 categories:

-Before steam generator replacements (SGR), -After SGR. The update of the data confirms that the Co-58 deposited activity is all the lower as the series is newer all along the period considered, 1450 MWe being the most recent series and 900 MWe the older one (except for the hot legs, where the gap seems to be reduced with time). Similarly, the contamination range narrowing with reactor types is confirmed when a longer period of measurements is available.

Several arguments have been proposed to explain this discrepancy among the 3 series: -pH300°C: the first reactors operated with a low pH, before implementing a boron/lithium coordination with a target pH300°C fixed at 7.2. But as mentioned in [START_REF] Anthoni | Effects of pH of Primary Coolant on PWR Contamination[END_REF][6] and reminded in [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], the impact of the pH on surface contamination is difficult to estimate from OPEX data analyses only. In paper [START_REF] Jobert | Study of the impact of three B/Li coordinated chemistries on the cobalt 58 surface contamination of a French PWR using the OSCAR code v1.4[END_REF], calculations have been performed with the OSCAR code to assess the effect of pH only on the surface contamination: the results showed a moderated impact of the pH, leading, for example, to a decrease of around 5 % of the Co-58 deposited activities in the SG when moving from a lithium content of 2.2 ppm to 3.5 ppm, in order to increase the pH300°C = 7.2 duration over the cycle. -SG tube alloys and SG tube manufacturing processes (see [START_REF] Dacquait | Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products[END_REF], [START_REF] Dacquait | The impact of steam generator replacement on PWR primary system contamination[END_REF]): The French PWR SG are equipped with tubing made of 600 Mill-Annealed (MA) alloy for the oldest, 600 Thermally-Treated (TT) or 690 TT for the most recent ones. The impact of the SG alloys can also be identified when considering EMECC measurements after SGR (see below: §2.1.2). In this paper, the impact of the manufacturing process of 690TT alloy SG tubes on the RCS surface contamination is assessed in §5.

For the primary pipes, whatever the series selected, the global trend of the curves is similar: after a sharp decrease of the Co-58 surface contamination within 6 to 8 cycles, followed by a smoother one, an almost stable state is reached. This evolution is consistent with the nickel release rate of SG tubes measured in loop tests. The plateau is reached sooner for the hot legs, compared to the cold legs. It can also be noticed that the dispersion of the values is globally decreasing given the cycles. This general trend is directly linked to the formation of the passive protective oxide layer on the surface of the SG tubes, decreasing the nickel release from the tubing over time. The replacement of the spacer grids, initially made in Inconel 718, by grids made in Zircaloy alloy, has also contributed to the reduction of the Co-58 surface contamination, as already explained in [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF]: the former spacer grids of the assemblies being composed mainly by nickel and submitted to the neutron flux, they likely constituted a non-negligible source of Co-58, whereas Ni can be present as impurity only in Zircaloymade materials.

For the SG, it is more difficult to clearly identify a trend (which tends to show that the evolution of the Co-58 SG contamination is highly unit-dependent), except for the newest series (1450 MWe), where a slight continuous decrease of the Co-58 surface contamination can be observed.

NPC 2023 -International Conference on Water Chemistry in Nuclear Reactor Systems France, Juan-les-Pins -2023, Sept. 25 -28

2.1.2. After SGR SG with 600 alloy tubes of the 900 MWe series have progressively been replaced by SG equipped with 690 alloy tubing since the 1990s. Figure 4 and Figure 5 present the mean Co-58 deposited activities after SGR, respectively in the HL, in the CL (or COL) and in the SG tubing of 900 MWe PWRs. Note that the first value displayed on each graphic corresponds to the state before SGR, in order to facilitate the comparison between the surface contamination before and after SGR. This single value before SGR represents the average of the contamination measured for the 6 cycles preceding the SGR. For primary pipes, after SGR, the global evolution of the Co-58 surface contamination is similar to that described in § 2.1.1, but the magnitude of the contamination is far much lower than observed with 600 alloy SG tubes and close to the level obtained with the most recent reactors (1450 MWe series). After 7 to 8 cycles, the Co-58 surface contamination is lower than the surface contamination state before SG. Although changes to limit the contamination have been introduced since the start-up of the 900 MWe plants, as mentioned above, the decrease of the Co-58 surface contamination is undoubtedly due to the new SG (decrease of the Ni content into the alloy and manufacturing process improvement -see [START_REF] Dacquait | Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products[END_REF] and [START_REF] Dacquait | The impact of steam generator replacement on PWR primary system contamination[END_REF]). Note that after the SGR, for the first cycles, the contamination is higher in the cold legs, compared to the hot legs, as it was before SGR.

Primary pipes:

For SG tubing, after their replacement, the Co-58 average surface contamination fluctuates for 6 to 12 cycles (as observed before SGR in §2.1.1 for the 900 and 1300 MWe series), depending on the SG subregions, with activities in the same order of magnitude as just before the SGR. Then, the Co-58 surface contamination reaches low values, similar to the 1450 MWe series which operates with 690 alloy tubing since the start-up.

2.2.

Co-60 contamination 2.2.1. Before SGR For the primary pipes, it is difficult to identify a common trend to the 3 series of the French fleet, for both Co-60 surface contamination of the hot legs and cold legs. On one hand, for hot legs, the 1450 MWe series shows a regular increase of the Co-60 deposited activities with time, which is not confirmed for the 2 other series (where the evolution of the contamination is more erratic, except for the first cycles where a sharp increase occurred).

On the other hand, for cold legs, an increase of the surface contamination for 4 to 10 cycles (depending on the series) is observed followed by a slow decrease over the cycles. Contrarily to the Co-58 situation (see §2.1.1), the discrepancy between reactor types is not as marked: the Co-60 surface contamination of the most recent reactors is not systematically lower than those of the 1300 MWe series. This observation is new and is in opposition of what could be expected from the first measurements on the 1450 MWe series (see [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF]).

As in the Co-58 case, the Co-60 contamination range narrows with PWR types, and the range of the Co-60 deposited activities is also decreasing with time, especially for the 900 MWe series.

For SG tubing, the measurements show a more homogenous trend: after an increase of the Co-60 surface contamination for 10 to 15 cycles (depending on the series), a stabilization trend is observed, and even a decrease in the 1450 MWe series case.

The SG tubings are all the more contaminated in Co-60 as the reactor series is old. Immediately after SGR, the Co-60 surface contamination is lower than the reference state before SGR in all parts of the RCS. Then, a steady increase of the contamination is measured:

After SGR

-Into the primary pipes, this increase is observed later in the cold legs than in the hot legs. In these two regions, the Co-60 deposited activities overpass that of before SGR: in the shortterm (4 cycles) for the hot legs, in the long-term (18 cycles) for the cold legs. -Into the SG tubing, this increase is slow and does not reach the reference level before SGR, even after 20 cycles.

Several origins of Co-60 surface contamination have already long been identified: -Wear of stellites surfaces used in valves, bearing and stops of pumps, control rod driven mechanisms (CRDM), core internals. The corrosion of stellites surfaces also contributes to Co-60 contamination but these releases are usually low compared to other Co-59 related sources, -Corrosion and release from alloys, especially from SG tubes (even if Co-59 is present as impurities in the alloy, the total wet surface of the tubing leads to a significant contribution to the Co-60 surface contamination of the circuits), -Release of activated in core materials, -Unplanned introduction of Co-59 in the primary coolant while undertaking maintenance operations during outage (for example, when grinding seat valves to avoid leakage), -Operating events (bearing degradation, for example).

As explained in [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], the answer to the question of the main contributor to Co-60 surface contamination depends on the available source term of Co-59 (surface of stellites, content of Co-59 impurities in the alloys ..) and its release rate (normal wear, unexpected events …), which is also linked to operating conditions (pH, dissolved H2 content …) or design options (CRDM load …). The sources being multiple, the main one depends on the level of the others.

When we compare the evolution of the Co-60 mean surface contamination before and after SGR on the 900 MWe series, along 30 cycles and 20 cycles respectively, it can be noticed that:

-For the primary pipes: Less variations of the values from one cycle to the next occurred after SGR than before; and when we limit the comparison of the Co-60 surface activities during the 15 first cycles of each period, the activity measured after SGR remains below that of before SGR. -For the SG tubing: 20 cycles after SGR, the Co-60 surface contamination is twice lower than the level measured just before SGR, in spite of the steady increase observed.

Regarding radiation field concerns, the SGR performed on the 900 MWe reactor series lead to a positive impact, by reducing the Co-60 surface contamination level in SG in a sustainable manner.

Prior to the SGR, (one of) the main source of Co-60 surface contamination in the 900 MWe series was probably the cobalt impurities of the SG in 600 alloy. Nevertheless, when confronting these Co-60 activities deposited in the tubing of the new 900 MWe SG to the others of the 1450 MWe SG (both are made in 690 TT alloy), they are still higher, around a factor 2 for a same cycle. That could reveal that one or several important contributors to the Co-60 surface contamination of the 900 MWe reactors equipped with 690 TT alloy SG tubing still exist, aside the question whether the new SG are still a major contributor or not.

Zinc injection

One of the key questions related to primary surface contamination in paper [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF] was the effect of zinc injection, which was about to be tested in 2004, and for which beneficial effects were expected, especially regarding Co-60 contamination. Since then, zinc injection has been widely implemented in reactors of the French fleet and a first analysis of the EMECC measurement campaigns conducted in France has been proposed [START_REF] Phlippot | 15 year experience of zinc injection at EDF[END_REF].

In PWRs, zinc is injected mainly to:

-reduce stress corrosion cracking, especially of 600 MA alloy (material aspects), -lower the radiocobalts source term, decrease the surface contamination in Co-58 and Co-60 and diminish the dose rates (source term and radiation protection aspects), and -limit and homogenise the deposit on the core and prevent axial offset anomaly (AOA) phenomena (fuel aspects). Nevertheless, the positive impact of zinc injection on radiation field is still on debate at the international level.

In this §, a more detailed analysis regarding the surface contamination effect of zinc injection is carried out from the updated EMECC data. Two kinds of zinc injection are distinguished:

-the preventive mode: zinc is injected continuously before SGR (preferably, injection must start at least 2 cycles before SGR) or at the early beginning of the cycle following the SGR, -the curative mode: zinc is injected after SGR (or after the start-up of a new reactor).

At EDF, zinc is currently injected in 23 reactors, belonging to the 3 series of the French fleet (900 MWe, 1300 MWe and 1450 MWe), either in preventive or curative mode.

The EMECC measures have been statistically studied, considering the preventive mode and curative mode separately. For each mode and each nuclide (Co-58 and Co-60), the following methodology has been applied: a comparative analysis of the evolution of the surface contamination between units injecting zinc and units without zinc injection is performed. The objective of this analysis is to identify whether units injecting zinc benefit from a lower surface contamination than units without zinc injection.

Figure 11 to Figure 14 present Co-58 and Co-60 surface activities, for different sub-regions of the RCS, for each injection mode. Data related to units without (respectively with) zinc injection are in blue (respectively orange).

 Preventive mode (900 MWe series): Figure 11 gives a general view of every Co-58 and Co-60 surface activities measurements in the RCS after oxygenation, given the number of cycles after SGR. This type of graphical representation can give the impression that units with zinc injection have a lower level of Co-58 surface contamination, and a medium to higher level of Co-60 contamination. Although, it might not be relevant to draw conclusions based on these figures only, since they do not allow to identify if one measurement (blue point or orange triangle) belongs to an upward or downward trend. Thus, they are completed by Figure 12, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to unit injecting zinc are displayed in colour (reactors B, C, D, E and F); those corresponding to units without zinc injection are in grey. Co-58 mean surface contamination after SGR tends to decrease in both cases, with and without zinc injection. But the decrease appears to be sharper when units are injecting zinc. When the analysis focuses on the different parts of the RCS, this positive impact is significant in cross-over legs and SG (cold part) (see §6). In the hot legs, a similar effect of zinc injection is not detected.

Regarding Co-60 mean surface contamination in the RCS after SGR, first a downward trend can be observed for units without zinc injection, sometimes followed by a slow upward trend. In contrast, for units injecting zinc, an increase trend occurs quickly and lead to higher Co-60 surface contamination in comparison to units without zinc injection. When the analysis focuses on the different parts of the RCS, this effect of zinc injection appears mainly into the hot legs (see § 6).

 Curative mode: Figure 13 presents the general aspect of Co-58 and Co-60 surface activities given the number of cycles. As previously, these are completed by Figure 14, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to units injecting zinc are displayed in colour; those corresponding to units without zinc injection are in grey. The beginning of zinc injection is indicated by red triangles (reminder: in curative mode, injections starts while the SG tubing surfaces are already passivated). For both Co-58 and Co-60 deposited activities, the trend of the coloured curves after zinc injection does not seem to differ to the one without zinc injection. But we can note that the period of time after zinc is injected is too short to provide a robust analysis. Moreover, when zinc starts to be injected, surface activities are already quite low, and a modification of the trend is more difficult to detect.

Recontamination during the cold shutdown procedure

Regarding source term management, the shutdown procedure of a reactor is of great concern, the risk of out-of-flux surface recontamination by corrosion products being one of the sensitive points. The shutdown procedure is characterized by a progressive decrease of the temperature and the pH, and the evolution of the redox potential (reducing to oxidizing conditions). The forced oxygenation implemented on the French fleet leads to the dissolution of corrosion products and the occurrence of activated corrosion products peaks, especially in dissolved Co-58. It is now well known that this Co-58 activity is due to the dissolution of metal nickel (and possibly of nickel oxide) in-core deposits, which leads to the dissolution of the Co-58 derived from the nickel sites in the crud ( [START_REF] Dacquait | Simulation of corrosion product transfer during PWR cold shutdown using the OSCAR code[END_REF]). Effective purification means (flow rate, cationic and/or anionic resins with the adequate chemical conditioning…) are required in order to avoid a recontamination (redeposition by precipitation) of the primary (and auxiliary) surfaces by activated corrosion products. To avoid transferring this considerable activity from in-core surfaces to out-of-flux surfaces, it is also required to start the oxygenation phase at a maximum temperature of 80 °C, and to reduce the duration between the chemical dehydrogenation and the oxygenation strictly speaking (when partial oxygenation is implemented with a chemical dehydrogenation). It has been shown, for example in [START_REF] Dacquait | Corrosion product transfer in French PWRs during shutdown[END_REF], that a recontamination of the primary circuit can occur if the oxygenation temperature of the primary coolant is higher than 80 °C. It is not excluded that Co-58 recontamination is accompanied by a cobalt and nickel redeposition. In such a situation, these elements are transferred and activated in the core during the following cycle, creating an additional source of radioactive corrosion products.

In this §, a first level analysis has been made for the 900 MWe and 1300 MWe series of the French fleet to assess whether the Co-58 and Co-60 recontamination risks during the oxygenation phase of the shutdown are well managed or not. Deviations under 20% are not considered significant, due to the measurement uncertainties. The root causes of the potential recontamination events observed remain to be identified.

Co-58 surface recontamination

Figure 15 and Figure 16 present the after to before ratios of Co-58 surface activities given the number of cycles, for the hot legs and the cold parts of the SG tube bundles respectively. Co-58 recontamination occurs in the HL and SG tubing (cold part) for both series. Nevertheless, for the 900 MWe series, the proportion of recontamination events seems to diminish with time. For the 1300 MWe series, the recontamination of the surfaces during oxygenation appears to be quite frequent into the SG tubing (cold part) sub-region.

Co-60 surface recontamination

Figure 17 and Figure 18 present the after to before ratios of Co-60 surface activities given the number of cycles, for the hot legs and the cold parts of the SG tube bundles respectively.

For Co-60 surface contamination, the ratios remain very close to 1, slightly below this value (decontamination trend). As function of cycles, no marked trend can be observed.

Manufacturing processes of SG tubes made in 690TT alloy

This study, as many others, outlines the great impact of SGR on Co-58 and Co-60 surface contamination of the RCS. Several papers based on operating experience data (for example [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], [START_REF] Dacquait | Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products[END_REF], [START_REF] Dacquait | The impact of steam generator replacement on PWR primary system contamination[END_REF]) highlighted the prominent role of the surface finish of the SG tubes related to their manufacturing process (more than the tube alloy) in the Ni-58 and Co-59 releases and thus the contamination of the circuits.

In this §, the objective is to check whether the manufacturing processes of a similar alloy is likely to also have an effect on surface contamination. Here, we considered EMECC data related to the 900 MWe series after SGR. All new SGs are equipped with 690 alloy, as mentioned previously, but different manufacturing processes have been used.

NPC 2023 -International Conference on Water Chemistry in Nuclear Reactor Systems France, Juan-les-Pins -2023, Sept. 25 -28

The impact of SG tube manufacturing processes and surface state have also been studied in papers using loop test experiments and characterization of both the internal surface of the tubes and the oxides formed (for example [START_REF] Carrette | Impact of the Surface State of Steam Generator Tubes on the Release of Corrosion Products in Pressurized Water Reactors[END_REF], [START_REF] Carrette | Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors[END_REF], [START_REF] Clauzel | Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant[END_REF]).

As explained in [START_REF] Carrette | Impact of the Surface State of Steam Generator Tubes on the Release of Corrosion Products in Pressurized Water Reactors[END_REF], since the first use of 690 alloy for SG tubes in France, the manufacturing processes have been modified: the sandblasting with corundum particles has been suppressed, the annealing initially realized under NH3 or H2 is only performed under H2 now, the carbon content has been reduced and the equivalent carbon content is taken into account for determining the annealing temperature, steps of pilgering and/or drawing have been optimized, cleanings have been introduced, cleaning baths have been modified etc.

For SG tubes equipping French reactors (with and without zinc injection), five main periods have been determined according to these changes: up to the end of 1988, 1989-1992, 1993-1995, 1995-2002, after 2002; the first one being the oldest and the fifth one the most recent. Each point of the following charts represents the average of the measurements gathered for the cycle considered, that is between 1 to 15 values by point. Given the low number of measurements for some points, the conclusion associated to this comparative analysis is preliminary.

Co-58 surface contamination

For each manufacturing process ("1" to "5"), Figure 19 and Figure 20 present the general aspect of Co-58 mean surface activities given the number of cycles after SGR. According to these data, for every region of the RCS, all manufacturing processes share a common global trend (sharp decrease after a few cycles followed by a slower one until a plateau after around 10 cycles). After 10 cycles, Co-58 surface contamination levels are very close (except, maybe, for the hot legs, where a discrepancy remains). The most recent manufacturing process (n°5) seems to present the best behavior (lower surface activity for the first cycles). Note that to strengthen this observation, the effect of the primary pipe Ni deposit level prior to SGR ("memory effect") should be factored in.

The oldest manufacturing process (n°1) presents high Co-58 surface contamination for the first cycles.

In the paper [START_REF] Carrette | Impact of the Surface State of Steam Generator Tubes on the Release of Corrosion Products in Pressurized Water Reactors[END_REF], this manufacturing process was also associated to a slower decrease of the Co-58 spike height. Although, there is no direct relation between the out-of-flux Co-58 surface activities and the Co-58 oxygenation peak (which is linked to the source term deposited on the fuel), this is representative of a significant release of nickel from SG tubing.

As the manufacturing process n°5 has been modified in comparison to the previous ones (evolution of the drawing step), it could be a hypothesis to explain the favorable behavior of these tubes.

Co-60 surface contamination

For each manufacturing process ("1" to "5"), Figure 21 and Figure 22 present the general aspect of Co-60 mean surface activities given the number of cycles after SGR. These data show that, for every region of the RCS, all manufacturing processes present a common global trend (steady slow increase of the Co-60 surface contamination; the beginning of a decrease for the last cycles seems nevertheless to appear for the manufacturing process n°3). The manufacturing process n°4 presents the best behavior, excepting at the beginning of the cycles in the hot legs, whereas the highest mean value on one cycle is reached for the manufacturing process n°5. However, it is impossible to conclude without a more comprehensive analysis, as many other factors influence the Co-60 surface activity evolution (stellites wear, cobalt impurities content in the SG tubes, contamination level of the pipes prior to SGR, unexpected material degradation, and maintenance operation, as mentioned previously).

Discussion

The previous EMECC measurement analysis [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF] outlined a "CP0 effect" among the reactors of the 900 MWe series. Indeed, before SGR, the primary circuit Co-60 contamination was significantly higher for CP0 units (the oldest French PWRs), compared to the CP1-CP2 ones, especially for the HL and CL (or cross-over legs) (less true for the SG tubing). This discrepancy among the reactors of the 900 MWe series is reminded in the Figure 23 below. After SGR, the updated data shows that the discrepancy between the 2 sub-series still remains on the primary coolant pipes but seems to disappear after around 10 cycles. In the SG tubing (see Figure 24), after SGR, the evolution of the Co-60 contamination is very similar when we compare the 2 subseries (CP0 vs CP1-CP2).

An unexpected increase of the post SGR Co-60 contamination in the legs of the CP1-CP2 reactors after 6 cycles has to be pointed out and the causes need to be investigated. In the HL, the average contamination reaches a level significantly above the one just before SGR and even higher to the Co-60 average contamination measured during the period with 600 MA alloy SG tubing. Due to the low number of measurements available from 15 cycles after SGR, it seems premature to conclude. Nevertheless, it appears as a weak signal.

In order to better characterize the different contributors to the Co-60 contamination, OSCAR calculations have been carried out and compared to EMECC measurements. The Figure 26 presents the evolution of the Co-60 contamination (for both measurements and calculation) before and after SGR of the reactor A, a 900 MWe series unit. Before SGR, these three sources contribute to the Co-60 contamination approximatively equally, with a slight preponderance of the stellites source, especially the one associated to the CRDM and the CVCS pump. Note that the Co-60 contamination associated to "in core materials and stainless-steel piping" is far from being negligible: it comes from the contribution of, before their replacement, the Inconel spacer grids, which Co-59 impurity content were around 800 Just before SGR, the Co-60 contamination is almost stabilized. After SGR, the Co-60 increase rates due to stellites and in core materials and steel piping are a continuation of what they were just before SGR and are thus very low in comparison to their levels at the start-up of the reactor A (stellites wear and corrosion of materials are maximum at the beginning of life).

With the new SGs, the combination of a corrosion/release rate significantly lower than the previous one (replacement of 600 alloy by 690TT alloy and evolution of the manufacturing process) and a reduction of the Co-59 impurity content into the 690TT alloy tubing contributes to the very low increase of the Co-60 contamination. Nevertheless, we cannot conclude that SG are the main contributors to Co-60 surface contamination, as the "production" of each source of Co-59 evolves with time. As outlined previously in [START_REF] Dacquait | Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products[END_REF] and [START_REF] Dacquait | The impact of steam generator replacement on PWR primary system contamination[END_REF], the Co-60 build-up coming from the remaining sources is very low compared to the former Co-60 contamination in the primary coolant pipes. According to usual trends of the SG corrosion/release rate, a stable level of Co-60 contamination or even a decrease within a few cycles can be expected (due to the radioactive decay; providing there is no detrimental operating event).

Zinc injection

The approach retained to analyze the possible effect of zinc injection on the surface contamination is to compare the evolution of the contamination of units with zinc injection to others without. In the case of the Co-60 contamination associated to the preventive mode injection, Figure 12 (right hand side) seems to show that the increase of the contamination after SGR is stronger and starts earlier for units B, C, D and E (which inject zinc) than for units without zinc injection. But data analysis can be tricky, and this preliminary conclusion needs to be strengthened by taking into account the radiological situation analysis of these units before SGR. Indeed, when considering the Co-60 deposited activities before SGR (and before zinc injection), some units may present lasting high contamination levels (resulting from stellites wear, operating events, maintenance operation…), temporarily lowered due to the SGR, which could also explain the increase trend after SGR. However, in the case of reactor B, the Co-60 surface activities evolution in the HL shows that high values have been reached (around 14 GBq/m 2 ) after 12 cycles but a decrease trend in the following cycles leads to lower values of 4 to 5 GBq/m 2 . Thus, it appears questionable to attribute the Co-60 surface contamination increase after SGR to the historical background of the plant before SGR.

Nevertheless, the finding that zinc injection could increase the Co-60 surface activities has to be moderated in the light of the following observation. Indeed, the global analysis focusing on average data (Figure 12, right hand side) coming from measurements in HL, CL, COL and SG may mask disparities in the evolution of the contamination depending on the sub-regions of the RCS considered. Figure 28 highlights this observation. Thus, in the SG, a positive effect of zinc injection regarding Co-60 surface contamination could begin to emerge, with a stabilization trend in the medium term of the deposited activities measured for reactors B, C, and D. Further data is needed to confirm this pattern. This increase of the Co-60 surface contamination could be linked to the antagonistic effect outlined in [START_REF] Bultot | Zinc effect on the primary circuit contamination of a Belgian PWR using the OSCAR V1.3 code[END_REF] regarding zinc injection: "when zinc is injected, Co dissolves from the inner oxide. A part of this surplus of Co is deposited on the fuel elements leading to an increase in the formation of Co-60. This activated deposit is eroded and the particles are deposited on the outer oxide of the out-of-core components, including the SG tubing, resulting in an increase in the Co-60 contamination of this outer oxide". Then, the effect of zinc on the corrosion, which lowers the release of Co-59 (passivation effect), could explain the stabilization of the surface contamination after the first cycles. To complete, we can mention here that reactors B, C, D and E are equipped with SG tubes manufactured with the process n°5 (the most recent one; see previously): although some reactors without zinc injection are also equipped with such SG tubes, given the observation presented in §5.1, this parameter may interfere in the analysis.

Conclusion

The analysis of the updated Co-58 and Co-60 surface activity data from the EMECC measurements campaigns provides the following lessons learned:

Regarding the global evolution of the contamination of the reactors of the French fleet:

-For Co-58: o The surface contamination values and the data scatter decrease with time as well as with the evolution of the reactor design (the most recent reactors are less contaminated than the older, for a same cycle), o In the 900 MWe series, SGR allows to further lower the contamination levels. After a few cycles (the number of which depends on the RCS sub-region), every part of the primary circuit reaches a surface activity below the one prior to SGR.

-For Co-60: o The positive effect expected of the newest reactor design is not so well marked, except in the SG sub-regions, o After SGR, the updated data for the 900 MWe confirm the smooth increase trend of the contamination in the SG but do not confirm in the longer term the decrease trend initially observed in the HL and CL. o After SGR, the analysis does confirm the "CP0 effect" outlined before SGR for the 900 MWe series but only for the mid-term: in the longer term, on the contrary, CP1-CP2 contamination levels tend to be higher than the CP0 ones and to exceed the values before SGR. As only a few data are available for the long-term period (> 15 cycles after SGR), it seems good to remain cautious in the interpretation of this observation.

Regarding the impact of zinc injection: -When zinc is injected prior to SGR (or at the very beginning of the cycle following the SGR), a favorable impact on Co-58 deposited activities is likely to be expected in the different parts of the RCS, -To be relevant, such an analysis needs:

o To follow the evolution of the contamination unit by unit, o To take into account the contamination historicity of each unit and especially the contamination levels prior to SGR, o To distinguish the data according to the location of the measurements (e.g.: HL, CL, COL, SG bundle…).

Regarding the impact on contamination of the manufacturing processes of the 690TT alloy SG tube bundles:

-the most recent manufacturing process seems to be associated to lower Co-58 surface activities, suggesting a positive impact of the manufacturing process evolutions introduced recently.

Last, this study points out the difficulties to assess, based on operating experience data, the impact of a single parameter on Co-58 and Co-60 surface contamination. Several parameters are likely to influence the surface contamination of the primary circuit, with possible antagonistic effects, and when comparing the deposited activities of two groups of reactors, the differences can rarely be attributed solely to the parameter of interest. In this respect, the development of data science-based analyses, which is planned in a near future, could lead to promising outcomes.

Figure 1 toFigure 1 :Figure 2 :Figure 3 :

 1123 Figure 1 to Figure 3 present the Co-58 deposited activities before SGR, respectively in the HL, in the CL (or COL) and on the SG tubing of French PWRs.

Figure 4 :

 4 Figure 4: After SGR -Evolution of Co-58 surface activities in HL (left) and CL&COL (right) of the 900 MWe series of the French fleet

Figure 5 :

 5 Figure 5: After SGR -Evolution of Co-58 surface activities in the SG tubing (hot and cold parts) of the 900 MWe series of the French fleet

Figure 6 to

 6 Figure 6 to Figure 8 present the Co-60 deposited activities before SGR, respectively in the HL, in the CL (or COL) and on the SG tubing of French PWRs.

Figure 6 :Figure 7 :Figure 8 :

 678 Figure 6: Before SGR -Evolution of Co-60 surface activities in the HL of the French fleet given the reactor series (maximum/minimum and average values)

Figure 9 andFigure 9 :Figure 10 :

 9910 Figure 9 and Figure 10 present the Co-60 deposited activities after SGR, respectively in the HL, in the CL (or COL) and in the SG tubing of 900 MWe PWRs. As in §2.1.2, the first value displayed on each graphic corresponds to the state before SGR (average of the contamination measured for the 6 cycles preceding the SGR).

Figure 11 :

 11 Figure 11: After 900 MWe SGR -Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection

Figure 12 :

 12 Figure 12: After 900 MWe SGR -Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)

Figure 13 :

 13 Figure 13: Before SGR -Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection

Figure 14 :

 14 Figure 14: Before SGR -Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)

Figure 15 :

 15 Figure 15: 900 MWe series -Evolution of Co-58 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)

Figure 17 :

 17 Figure 17: 900 MWe series -Evolution of Co-60 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)

Figure 19 :

 19 Figure 19: 900 MWe series -Evolution of Co-58 surface activities in the HL (left) and CL&COL (right) given the manufacturing process

Figure 21 :

 21 Figure 21: 900 MWe series -Evolution of Co-60 surface activities in the HL (left) and CL&COL (right) given the manufacturing process

Figure 23 :

 23 Figure 23: 900 MWe -Comparison of Co-60 surface activities in the HL (left) and CL&COL (right) between the CP0 and CP1-CP2 sub-series before SGR (minimum/maximum and average values)

Figure 26 :

 26 Figure 26: OSCAR calculation and EMECC measurements -Evolution of Co-60 surface contamination before and after SGR in the HL of the reactor A (900 MWe series)

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Figure 27 :

 27 Figure 27: Reactor B -Evolution of Co-60 surface contamination in HL before SGR

Figure 29 :

 29 Figure 29: After SGR -Evolution of Co-58 surface activities in the SG with and without zinc injection