J B Génin

Hervé Marteau

Frédéric Dacquait

Gilles Benier

Jerome Francescatto

Francois Broutin

Frédéric Nguyen

Mariane Girard

G Bénier

L Noirot

S Maillard

V Marelle

A Bouloré

D You

G Plancque

G Ranchoux

J Bonnefon

V Bonelli

M Bachet

G Riot

F Grangeon

The OSCAR code package : A unique tool for simulating PWR contamination

ou non, émanant des établissements d'enseignement et de recherche français ou étrangers, des laboratoires publics ou privés.

INTRODUCTION

The assessment of the radio contamination of PWR primary and auxiliary circuits is a crucial issue for power plant operators and suppliers. It is necessary to fulfill some important challenges while operating or designing a nuclear power plant : decreasing the impact on personnel exposure to radiation, optimizing the plant operation, limiting the activity of the wastes produced during the reactor lifetime and preparing decommissioning.

The radio contamination of PWR primary circuits comes from corrosion products, fission products and actinides. Their origin is different but they are subject to the same severe conditions (300 °C, 150 bar, neutron flux, water velocity up to 15 m.s -1 Moreover during a reactor cycle, several kilograms of corrosion products (mainly nickel and iron) are released in the primary coolant from the base metal and represent most of the contamination mass deposited on the primary surfaces. So their behavior definitely determines the evolution of deposited radioisotopes : radioactive corrosion products of course, but also actinides.

) and follow the same transport mechanisms in the primary circuit. The main processes involved are erosion/deposition, dissolution/precipitation, adsorption/desorption, convection, purification, neutron activation, radioactive decrease.

Therefore, the two existing French contamination codes (PACTOLE [START_REF] Dacquait | Simulations of Corrosion Product Transfer with the PACTOLE V3.2 Code[END_REF] for corrosion products, and PROFIP [2] for fission products and actinides) have been merged in a unique tool : the OSCAR code package, dedicated to the PWR primary and auxiliary circuit contamination. This paper describes the features of this new code package, and first results of validation against experimental data.

After a general overview of the new features of the code in section 2, the source terms in the primary circuit are described in section 3 for the different radioisotopes considered. Section 4 lists the main mechanisms involved in the transport of the radioisotopes and the contamination of the primary circuit. Finally in section 5, the first validation of the code against experimental data is shown for the main isotopes.

Tramp uranium

FP recoil emission

Fuel oxide dissemination Fission Products release

Figure 1. Fission products and actinides release 2. GENERAL OVERVIEW

The OSCAR code has been developed within the framework of R&D cooperation between the French Atomic and Alternative Energy Commission (CEA), Electricité de France (EdF) and AREVA NP.

The OSCAR code keeps the fully object-oriented design of PACTOLE V3 and its modular structure, as well as its discretisation into control volumes. A user friendly interface for input data and visualisation of results has been developed. In the version 1.1 of the OSCAR code, the calculation for fission products and actinides on the one hand, and corrosion products on the other hand, are still independent though they can follow the same mechanisms, as for neutron activation and radioactive decay.

Concerning the transfer of corrosion products, the PACTOLE V3 modeling [START_REF] Dacquait | Simulations of Corrosion Product Transfer with the PACTOLE V3.2 Code[END_REF] is retained, with an improved thermodynamic database. Unknown input parameters have been assessed using a multi-criteria calibration based on a comparison with experimental data (see section 5).

Compared to the PROFIP code, the main change in the modeling of the transfer of fission products and actinides is the use of the fuel code ALCYONE and the fission gas behavior module MARGARET [START_REF] Noirot | MARGARET : An advanced mechanistic model of fission gas behavior in nuclear fuel[END_REF], to assess the fission product source term in the primary circuit, released from a defective rod. Mechanisms involved in the primary circuit behavior are the same as for the PROFIP code. However, a model for adsorption/desorption has been added to describe the strontium behavior.

THE SOURCE TERMS

In this section, the source terms taken into account in the OSCAR V1.1 code are presented for the different types of contamination products : fission products and actinides released from defective fuel rods, fission products emitted by recoil from actinides deposited on surfaces under neutron flux (see fig. 1), and corrosion products released from the base metal.

Fission Products released from defective fuel rods

In case of rod cladding defect, fission products are released in the primary circuit from the fuel rod. Three stages can be distinguished in this process : release out of the rod pellets, transfer through the fuel-cladding gap and release through the defect.

3.1.1

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

Release out of rod pellets MARGARET [START_REF] Noirot | MARGARET : An advanced mechanistic model of fission gas behavior in nuclear fuel[END_REF] is an advanced mechanistic model of fission gas behavior in PWR fuel. It is developed by CEA within the NPC 2010 October 3-7, 2010 framework of the PLEIADES [START_REF] Michel | PLEIADES: multi concept nuclear fuel modeling[END_REF] platform and its fuel behaviour application ALCYONE, which supplies the thermo-mechanical data used by MARGARET.

The MARGARET module, which initially dealt only with stable gas, has first been adapted to the radioactive fission products, with validation against the experimental irradiation campaigns HATAC and CONTACT (fission product release measurements from irradiated tight fuel rods).

It has been shown that the bubble trapping of fission products inside the fuel grain could be neglected for fission products with radioactive half-life shorter than 8 days. These short life isotopes are really the isotopes of interest because they represent most of the primary fluid activity. Moreover, they are used in the criteria defined in the French PWR operating specifications, allowing or not the operation to be continued in the presence of cladding defects. In particular, the percolation mechanism is not taken into account. This implies that in this situation, radioactive fission products at grain boundaries are never released in free volumes and that the release from the fuel pellet into the free volume can be expressed as the result of pure diffusion in the fuel grain, towards free surfaces (cf. figure 2). Fission product concentration at the free surface is considered to be zero.

I.

Thus, as intragranular bubbles do not interact with short life fission products, the intrinsic diffusion coefficient of stable gas in the fuel oxide crystal is used for these isotopes. Like the stable gas, radioactive fission products are released from the grain by diffusion, either towards grain boundaries or towards free surfaces, that means grain surfaces connected with free volumes (fracture or porosity or fuel-cladding gap). In the former case, they may migrate towards the free volumes by percolation, when the bubbles at the grain boundaries interconnect. The free surfaces fraction is simulated by the MARGARET module, and is around 1 or 2 % in a tight fuel configuration.

The released fraction in free volumes can then be expressed as : D : diffusion coefficient of the fission products in the grain λ : radioactive decay constant of the fission product.

λ D a F B R free ⋅ ⋅ = 3
For defective rods, oxidation of the fuel matrix leads to released fractions much higher (by a factor of roughly 30) than for non defective fuel rods. This impact from oxidation is taken into account in the simulation by the following adjustments :

• Fraction of free surface F free

• Diffusion coefficient inside the grain D : a correction factor is applied to the intrinsic stable gas diffusion coefficient used for tight fuel.

: it is fixed to 100%, which prevents percolation from playing any role.

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

The calibration of the diffusion coefficient and of the thermal conductivity degradation is performed against a unique experimental database including defective PWR rods well identified and characterized, as well as analytical irradiations in the SILOE reactor. The release rates are deduced from primary activity measurements in fission products which are linked with the release from the fuel matrix and not affected by the transfer in the fuel-cladding gap and through the defect. These fission products are 133 Xe (if the burnup is low enough to allow an open gap) and 131 I (the release is correctly assessed from transient measurements). The fission products with a shorter life time are considered to be too strongly impacted by their transfer in the gap.

NPC 2010 October 3-7, 2010 Figure 3 gives a comparison between measurements and OSCAR calculations after calibration, concerning the released fractions from defective UO 2 fuel rods of the database (including EDITH 02 and EDITH 06 analytical irradiations, respectively for low and high burnup UO 2 3.1.2 rods). The validation domain is up to 50 GWd/t. We note a relatively good agreement between experimental and calculated released fractions. Such a fair agreement (not shown here) is also found for analytical irradiations EDITHMOX 01 (low burnup MOX rod) and EDITHMOX 02 (50 GWd/t MOX rod), which allows to widen the validation domain to MOX fuel, for the same burnup range.

To fit the experimental release of short life products at low temperature, it is necessary to take into account the knockout mechanism at the grain free surface [5]. That means ejection of an atom struck by a fission fragment or another atom of the lattice, due to elastic collisions. The important parameter for this mechanism is the thickness of the layer where knockout can lead to the ejection of fission products out of the grain. This thickness is assessed about 5 nm.

Knockout mechanism

0.00E+00 The knockout flux can be expressed as an additional sink term in the knockout layer to the diffusion term. This flux is proportional to the fission rate and to the fission product concentration in the layer. ), which are used for defective fuel calibration but strongly affects shorter life time fission products, particularly at low fuel temperature.

Release out of the rod represents the main source term for short life fission products in the primary circuit.

Release out of the rod

For the noble gases of interest, as long as the gap is not closed, the transit in the fuel-cladding gap is fast and the release out of the rod corresponds to the release out of the fuel matrix (unless the defect size is very low).

Considering iodine isotopes, the release mechanisms are not well understood. During nominal reactor operation, a trapping occurs in free volumes, which makes the release rate out of the rod very difficult to assess. This trapping phenomenon is taken into account in the OSCAR code through a trapping coefficient. However, during power transients at low level, water enters the free volumes and leads to the release of almost the total amount of iodine present in free volumes. The OSCAR code can be used to assess the iodine amount which is trapped in free volumes and released during a power transient. However, it can not reproduce the kinetic of the iodine release which is observed during the transient.

Fuel oxide dissemination

In the event of a significant cladding damage, fuel oxide may disseminate into a PWR primary circuit. The disseminated fuel oxide is also called "tramp uranium". No phenomenological modeling is linked with this dissemination. However experimentally, the dissemination rate can be assessed from the 134 Thus, the rate and duration of the fuel oxide dissemination are fixed in the data set of the OSCAR code, in order to reproduce the experimental I primary volume activity measurement, which is a tracer of fissile material deposited under neutron flux (cf. § 3.3).

134

The fuel oxide dissemination is the source term for actinides in the primary circuit. It is also a significant contribution to the source term of long life fission products (such as I activity level during the dissemination reactor cycle. The isotopic composition of fuel oxide is the same as the one at the fuel pellet surface, and is calculated by the ALCYONE code. 90

Fission Products emitted by recoil from the tramp uranium

Sr), which are included in the fuel oxide.

When deposited under neutron flux, tramp uranium generates fission products, which are released in the primary circuit by the recoil mechanism.

Unlike the diffusion mechanism which is involved in the fission product release out of a defective rod, the recoil mechanism is immediate, which facilitates the release of short life fission products such as 134 Recoil from the tramp uranium is another source term for the noble gas and iodine isotopes listed in § 3.1. This is the main source term for fission products which are not released from the fuel oxide inside the rod (such as I (radioactive half-life : 54 min).

90

A reliable assessment of this source term requires a correct calculation of the tramp uranium evolution in the primary circuit, which is mainly due to the neutron flux and to the transfer through erosion/deposition mechanisms between the different surfaces. Then the release by recoil is easily obtained by means of fission cross sections, which are specially assessed for out of pellet fissile material. Sr) or which are trapped in free volumes (such as iodine).

NPC 2010

October 3-7, 2010

Corrosion Products

An empirical approach for corrosion and ion release processes has been chosen, as for the PACTOLE code modeling [START_REF] Dacquait | Simulations of Corrosion Product Transfer with the PACTOLE V3.2 Code[END_REF][START_REF] Nguyen | First numerical simulations of contamination of the PWR primary circuit by activated corrosion products with the PACTOLE V3.0 code[END_REF] (see figure 5).

Uniform and generalized corrosion of the base metal (mainly stainless steels and alloys containing nickel or iron) leads to the formation of an oxide layer (with a given corrosion rate) and to the direct release of ions in the primary coolant (with a given release rate). In fact, the oxide layer is composed of an inner layer, which limits the release of ions in the primary coolant but without preventing it completely, and of an outer layer, which may dissolve in the primary fluid, and so contribute to the source term.

The corrosion rate and the ion release rate are determined on the basis of test loop experiments [7]. They depend on the physico-chemical conditions in the primary circuit (pH, temperature, boron and hydrogen concentrations) and on the components (surface treatment, manufacturing process, exposure time in the circuit). Moreover an attenuation factor is applied on the release rate when the equilibrium concentration is approached in the primary fluid. This is done in order to take into account the fact that the release of corrosion products is mainly a dissolution mechanism, which occurs only when the solubility limit is not reached.

In the model, the release of corrosion products in the primary circuit corresponds to several kilograms of iron and nickel, which determine the behavior of the deposited contamination, including actinides (a few grams) and insoluble fission products (a few milligrams).

The released corrosion products are transported by the coolant to the core where they can deposit on fuel assemblies. Such corrosion product deposits can be activated and subsequently set back into the coolant where they can redeposit on out-of-core surfaces and raise radiation fields. Moreover, in the case of surfaces under neutron flux, the activation mechanism (see § 4.1) generates radioactive corrosion products in the base metal, which are directly released in the primary circuit through the release process.

MECHANISMS WHICH GOVERN THE CONTAMINATION BEHAVIOR

This section presents the main physico-chemical mechanisms involved in the contamination behavior in the primary circuit for the OSCAR V1.1 version. They generate transfers between the following media : ions, particles and deposit. For each of these media (as well as for metal and oxides), for each control volume and for each isotope considered, a mass balance is calculated.

Together with the source terms, these mechanisms are summarized in figure 6 for actinides and fission products and in figure 7 for corrosion products. Besides the mechanisms described below, the OSCAR code considers the convection, which transports ions and particles through the primary circuit, and the purification, which removes ions and particles from the circuit with a given efficiency when they go through the filter or demineralizer of the purification circuit.

Neutron activation, fission, radioactive decay

The considered isotopes are described within filiation chains, with all required data to define the reaction between two isotopes : radioactive decay constant, branching ratio, reaction nature, cross section and fission yield if necessary.

Radioactive decay and neutron capture reactions (including fission reactions) occur either inside the materials under flux, generating radioactive isotopes inside the base metal (see § 3.4), or in the deposited contamination or in the fluid. For each capture reaction (between two isotopes) and for each location (surface or fluid), reaction rate libraries (for corrosion products) or one energy group cross section libraries (for fission products and actinides) are used, obtained from APOLLO calculations and JEF 2.

F u e l o x i d e d i s s e m i n a t i o n

or MOX), the initial enrichment and the burnup. For example, these mechanisms allow us to simulate the tramp uranium evolution when deposited under neutron flux, as well as the fission products creation from the fissile nuclides deposited under flux or the 58

Dissolution/precipitation

Co creation inside the nickel deposited on the core.

Precipitation or dissolution takes place, depending on whether the volume concentration of one element is higher or lower than the equilibrium concentration respectively in the considered control volume.The equilibrium concentration is calculated using the PHREEQCEA code (a modified version of PHREEQC [START_REF] Parkhurst | User's guide to Phreeqc (version 2) : a computer program for speciation, batch-reaction, one-dimensional transport, and inverse geochemical calculations[END_REF]), in combination with a thermodynamic database developed by CEA. It depends on the coolant temperature in the control volume as well as boric acid, lithium, hydrogen and oxygen concentrations, but also on the composition of the outer oxide and the deposit. Indeed these two media, as well as the ions and particles, are considered for the thermodynamic equilibrium calculation.

Dissolution and precipitation induce transfers between deposit and ions in the control volume. Dissolution kinetic is calibrated from the experimental data (see § 5.4). As for the kinetic of transfer to the wall due to the precipitation process, it is supposed to be limited only by the mass transfer coefficient between the fluid and the wall, and depends on the thermo-hydraulic conditions. It is much higher than the dissolution kinetic.

In the OSCAR 1.1 version, these two mechanisms concern only corrosion products. For example, they allow us to simulate that the nickel element, which is mainly released from the Inconel TM Precipitation of ions on particles is also taken into account, but does not seem to have a significant influence on the contamination behaviour with OSCAR V1.1 and is not represented in figure 7.

steam generators, precipitates in the hot areas of the core. Indeed the nickel solubility calculated using the CEA thermodynamic database is lower at the core exit temperature than at the entrance temperature.

Erosion/deposition

Erosion and deposition mechanisms concern both corrosion products and actinides.

The modeling assumes that the deposit erosion occurs only when the deposit thickness has reached a limit thickness, which depends on the thermohydraulic conditions and on the surface roughness. This mechanism generates particles in the primary fluid, with the same composition as for the deposit.

The erosion rate is calibrated from experimental data separately for actinides on the one hand and for corrosion products on the other hand. Moreover, the limit thickness considered for actinides is supposed to be zero, so that erosion occurs as soon as actinides are deposited. On the contrary for corrosion products, only surfaces for which the deposit has reached the limit thickness are concerned by the erosion mechanism.

The deposition model induces a particle transfer to the deposit. The corrosion product deposition rate is supposed to be mainly due to the turbulent diffusion and the effects of inertia [START_REF] Beal | Deposition of particles in turbulent flow on channel or pipe walls[END_REF] and depends on the thermohydraulic conditions in the control volume, and on the particle size. A log normal distribution centred on 0.5 µm is considered for the particle size distribution. Concerning the actinides deposition rate, it is calibrated from the experimental data.

NPC 2010 October 3-7, 2010

For example, the nickel deposit (due to precipitation) on the top part of the core reaches the limit thickness and the erosion process generates particles rich in 58 The erosion/deposition process is also responsible for actinides transfer between steam generators and the core, generating recoil emission of fission products, even after the contaminated core has been discharged.

Co (nickel activation product). Then these particles contaminate (through the deposition process) the out-of-core surfaces.

Adsorption/desorption

In the primary circuit conditions, strontium is found as Sr 2+ ion in the fluid. However important amounts are present on the primary surfaces, which can be explained by the adsorption process.

Therefore, adsorption and desorption mechanisms have been introduced in the OSCAR code to simulate the 90 Sr behavior in the primary circuit ( 90The mechanism is determined by the equilibrium ratio between the activity in the fluid and the activity on the surfaces, which depends on control volume temperature and pH. During nominal reactor operation, calculated surface and volume activity of Sr radioactive half-life : 29 years).

90

During an end-of-cycle shutdown, the decrease of pH and temperature leads to a significant decrease of the modeled equilibrium ratio, and then to a strong release of Sr reach an equilibrium : the volume activity depends only on the source term and the purification term, as surface activity is given by the equilibrium ratio. 90

EXPERIMENTAL VALIDATION FOR SPECIES HANDLED BY OSCAR

Sr in the primary fluid.

Considering the source terms and mechanisms described in sections 3 and 4, OSCAR is able to simulate the behaviour of four groups of contaminant species : actinides, gaseous and volatiles fission products, 90 In the OSCAR 1.1 version, actinides and fission products simulation is still independent from corrosion products simulation.

Sr and lastly corrosion products. As it appears in figure 6, in case of the presence of tramp uranium consecutive to fuel oxide dissemination, actinide contamination is a source term for fission products and must be calculated prior to fission product contamination.

We give here some examples of comparison with measurements, obtained from primary fluid analyses or measurements campaigns with EMECC device, a special gamma spectrometry device allowing surface activity measurements [START_REF] Eimecke | Ensemble de mesure et d'étude de la contamination des circuits (EMECC)[END_REF].

Xe, Kr, I

Gaseous and volatile fission products are totally soluble and do not interact with the primary walls in the nominal operating conditions. The Henry's law coefficient is used to evaluate the partition between liquid and gaseous phases, and to calculate the activity in the gaseous phases of the purification tank and of the pressurizer.

In figure 8, we show a comparison between measurements and OSCAR calculations for 133 The main uncertainty for fission products calculation is the source term coming from the defective rod. This uncertainty is important for iodine, because the trapping in the fuel-cladding Xe primary volume activity during a PWR cycle with occurrence of a cladding defect on a first cycle rod. A fair agreement between measurements and calculations is found. gap is difficult to model. Concerning noble gases, the uncertainty is low as far as the fuelcladding gap is open. As soon as the gap gets closed, a mechanical retention may occur in the gap, which is not modeled.

Actinides

The actinides are considered to be totally insoluble. Their evolution in the primary circuit is due to the neutron capture reactions and to the transfers determined by the erosion/deposition process. The erosion and deposition rates are difficult to evaluate and this results in the main uncertainty for simulation of actinides distribution in the primary circuit.

Concerning actinides, OSCAR calculations are very close to PROFIP calculations, which have been presented in [2] and are not shown here.

Strontium

Strontium is considered to be found as Sr 2+

Figure 9 gives a comparison between measurements and calculation during a refueling shutdown. It shows that the experimental peak of ion in the primary circuit, and follows the adsorption/desorption process. The source term comes from recoil emission from deposited tramp uranium under neutron flux and in case of fuel oxide dissemination, from strontium included in the fuel oxide. 90 Sr primary activity at the end of a cycle is correctly reproduced by the desorption model (effect of the decrease of temperature and pH) and by the purification model, which removes 90 The main mechanisms involved in corrosion product transfer are dissolution/precipitation and erosion/deposition. As already mentionned, the modeling is the same as for PACTOLE V3 [START_REF] Dacquait | Simulations of Corrosion Product Transfer with the PACTOLE V3.2 Code[END_REF], except that the thermodynamical database has been improved. Moreover, a multicriteria calibration has been performed to calibrate the parameters which are not well known. The parameters to be calibrated have been chosen amongst the most influent. They are the following : erosion rate, dissolution kinetic, metal release (within the range fixed by the experimental results), initial deposit mass (thought to be due to the hot functional tests prior to the first cycle). Besides, an experimental program is under progress to determine the oxide dissolution kinetic [11].

Once calibrated, the OSCAR calculations are compared for the main radioisotopes ( 58 Co, 60 Co) to the deposited activities on hot legs, cold legs, steam generator tubes, as well as to the volume activities. As illustrated in figure 10

CONCLUSION AND PERSPECTIVES

The OSCAR code package is a new step in the simulation of PWR contamination. With the same code, it is now possible to simulate the contamination of corrosion products, as well as fission products and actinides. Nevertheless, the qualification of this new version needs to be extended on more experimental results (reactor cycles).

The OSCAR code package results from the merging of two codes : PACTOLE and PROFIP. Compared to these previous codes, significant improvements are obtained with the use of the ALCYONE/MARGARET code for defective fuel rod calculations, as well as the use of an improved thermodynamical database and a multicriteria calibration for corrosion products. Moreover, a model for 90 However, in this first version of OSCAR, the fission products and actinides calculations are still independent from corrosion products calculations. In the next version, we will consider the transport of actinides and other non soluble fission products by the corrosion products, so that the erosion and deposition mechanisms will be determined by the corrosion products, even for actinides and fission products. This will allow to describe simultaneously in a consistent framework the coupled behavior of corrosion products and fission products/actinides. Sr behavior has been added.

Besides, we will consider a soluble form of actinides, specially after the primary fluid oxygenation at the end of the cycle, and the setback of actinides in water due the corrosion product dissolution. These new models will allow the simulation of alpha activity in the primary fluid during a refuelling shutdown. Future versions of OSCAR will also be able to implement models for other isotopes as soon as they become available.

Figure 3 .

 3 Figure 3. Comparison between experimental and calculated released fractionsfor UO 2 defective fuel.

Figure 4 .

 4 Figure 4. Effect of the knockout model.

Figure 4

 4 Figure 4 illustrates the impact of this mechanism, showing the variation of the release rate for different values of the diffusion coefficient. The doted lines represent the asymptotes for the pure diffusion mechanism and the red lines the diffusion + knockout mechanisms. As seen in this figure, knock out does not affect significantly released fraction for 133 Xe (λ = 1.53 10 -6 s -1 ) and 131 I (λ = 1.00 10 -6 s -1

FigureFigure 6 .

 6 Figure 5. Media taken into account in PACTOLE code modeling [1].

Figure 7 .

 7 Figure 7. Main mechanisms involved in the corrosion products contamination.

  2 nuclear database. In these libraries the values depend on the reactor type, the fuel nature (UO 2

Figure 8 .

 8 Figure 8. Calculation/Measurement comparison for 133 Xe primary volume activity after cladding defect occurrence.

  Sr ions from the primary fluid.

  Figure 9. Calculation/Measurement comparison for 90

Figure 10 .

 10 Figure 10. Activity deposited inside the SG tubes.

  These isotopes of interest are : 133 Xe, 133m Xe, 135 Xe, 138 Xe, 85m Kr, 87 Kr, 88 Kr, 131 I, 132 I, 133 I, 134 I,

	135
	Considering the defective fuel
	rod situation, experimental data
	show that the fuel burnup has
	no impact on the radioactive
	fission products release from
	the fuel pellet. Hence the
	mechanisms which induce an
	enhancement of the release of
	stable gas with the burnup are
	not considered for defective
	fuel.

GRAIN FREE SURFACE FREE VOLUME GRAIN BOUNDARY Pure diffusion trapping Figure 2. Radioactive fission product release out of fuel oxide grain in defective fuel rod configuration.

	NPC 2010	October 3-7, 2010
	R: release rate from the grain into free volumes
	B : birth rate in the grain
	F free	: fraction of grain surfaces connected with free volumes
	a : radius of the grain


G Ranchoux

F Dacquait

E Velasco-Marcelo

J M Garces De Marcilla

Bayo Cnat

/ Trillo

EMECC CAMPAIGN AT TRILLO : RESULTS AND COMPARISON WITH FRENCH FEEDBACK

The reduction of collective dose is an important objective for all nuclear plant operators. One of the support for studying the impact of design and operation parameters is the measurement campaigns of contamination in PWRs. EDF performs consequently gamma spectrometry measurements with CEA and his EMECC system, with more than 300 campaigns in EDF plants and also some campaigns in collaboration with other operators.

Commissioned in 1988, the only unit of the nuclear power plant of TRILLO, located in Spain, is a 3-loop PWR of 1066 MWe designed by KWU-Siemens. TRILLO-1, equipped with alloy 800 tubed steam generators, is characterized by a low dosimetry. In order to compare the contamination level of TRILLO-1 to that of French PWRs, an EMECC campaign was performed within the framework of a contract between EDF and the CEA, and with the agreement of CNAT/TRILLO and the support of the CSN (Consejo de Seguridad Nuclear). This campaign took place during the refueling shutdown at the end of cycle 21 from 9 to 16 February 2009. A previous EMECC campaign was performed in 1992 during the refueling shutdown of cycle 4.

At the first time, this paper will present the main conclusions of this EMECC campaign :

• deposited activities measured on legs walls and steam generator tubes walls,

• dose rates measurements.

The evolution of the contamination since cycle 4 will be also analyzed.

Furthermore, a comparison with the French unit St-LAURENT B1 will be achieved in order to better understand the origin of the differences (source term or outage management) and to improve good operational practices in the future.

1.

INTRODUCTION

Commissioned in 1988, the only unit of the nuclear power plant of TRILLO, located in Spain, is a 3-loop PWR of 1066 MWe designed by KWU-Siemens. TRILLO-1 equipped with alloy 800 tubed steam generators, is characterized by a low dosimetry (382 Man.mSv in 2008).

In order to compare the contamination level of TRILLO-1 to that of the French unit St-LAURENT B1, an EMECC campaign [1.] was carried out within the framework of a contract between EDF and CEA and with the agreement of CNAT/TRILLO and the support of the CSN (Consejo de Securidad Nuclear i.e. the Spanish Nuclear Safety Authority). This campaign took place during the refueling shutdown at the end of cycle 21 from February 9 th to 16 th

OPERATIONNAL PARAMETERS AT TRILLO

2009. A previous EMECC campaign was conducted in 1992 during the refueling shutdown of cycle 4.

Reactor coolant chemistry

In TRILLO-1 during cycle 21 (281 EFPD), the reactor coolant was based on a boron-lithium coordinated chemistry with a pH 300°C target of 7.5. Almost immediately at the beginning of cycle 21, pH 300°C rose above 6.9 whereas the maximum Li content has remained quasi-stable at about 2.2 ppm for 80% of the cycle duration. This procedure is slightly different than the French one for which the pH 300°C target is 7.2 (optimum chosen by EDF between contamination risk and potential PWSCC long term risk with high lithium concentration).

During cycle 21, H 2 concentration in primary coolant ranged between 22 and 28 cm 3 /kg (TPS), with an average of 24 cm 3

EMECC CAMPAIGN RESULTS AT TRILLO-1 AND COMPARISON WITH SAINT-LAURENT B1

Volume activities

During cycle 21

A comparison of volume activities during TRILLO 1 cycle 21 and St LAURENT B1 is shown in table 1. One can observe that the volume activities in CNT121 are slightly lower than those measured in SLB1.

During shutdown

RCS oxygenation is the most important stage of the French cold shutdown procedure because it usually causes a very large increase in the 58 [8.] Co volume activity (order of magnitude of 100 GBq/t), which can lead to an increase in the activities deposited on the out of flux surfaces . To control this stage, H 2 O 2 is injected when the primary coolant temperature is 80°C. Nevertheless, due to an improved manufacturing process of the alloy 690 TT SG tubes (Ni : 60%; Fe : 10%; Cr : 30%) of St-LAURENT B1 [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], the 58 Co peak is very low at cycle 20 : 0.8 GBq/t.

In TRILLO 1, thanks to the alloy 800 SG tubes (Ni : 35%; Fe : 45%; Cr : 20%), the primary system oxygenation leads to a very low 58 Co peak (order of magnitude of 1 GBq/t) typical of Konvoy PWRs, therefore there is less need to control it. At cycle 21, the primary system oxygenation occurred during the start of RHRS, perhaps during the draining of RCS (increase of 110m Ag volume activities) and of course during the primary system ventilation to atmosphere.

The 58 Co peaks in PWRs equipped with alloy 690 SG tubing, like St-LAURENT B1, can be as low as those of the Konvoy PWRs thanks to a very low SG tubing release. Therefore, more than the material (alloy 690 or alloy 800), the key factor leading to very low 58 Moreover, in TRILLO 1, the low Co peaks is the SG tubing surface state obtained by means of an improved tubing manufactured process.

58

Co peak could be explained by the fact that practically all the nickel released by the alloy 800 SG tubes generates nickel ferrites. Thus, there is almost no Ni 0 or NiO deposited on the fuel elements and potentially releasable during the oxygenation peak. Nickel ferrites are more stable in oxidizing medium and their solubility increases with temperature in optimum conditions (no preferential deposits on fuel elements). In St-LAURENT B1, the lack of Ni 0 [7.] or NiO is due to a very low release of alloy 690 SG tubing .

Surface activities

Average surface activities measured at TRILLO 1 cycle 21 are presented in table 2 for legs, SG tubes and CVCS non regenerative heat exchanger (NRHE). , can be explained by thermochemical considerations. Silver is soluble in oxidizing conditions and its solubility increases with temperature. Thus, during oxygenation, it dissolves in the primary coolant from RCS deposits and then precipitates on the cold parts of auxiliary systems, in particular heat exchangers.

Deposited activities in 58 Co and 60 Co are presented in figure 2 for hot legs, in figure 3 for crossover legs, in figure 4 for SG tubes and in figure 5 for CVCS non-regenerative heat exchanger. There is no difference in contamination level between SG tubes hot and cold sides at TRILLO 1 whereas the SG tube cold side at St LAURENT is about 2 times higher than the hot side. This phenomenon has rarely been pointed out on the French fleet and has been mainly observed after a steam generator replacement [3.]. This could be explained by a probable strong dependence of the new SG tube oxides solubility to temperature due to differences into stoechiometric ratios between old and new SG tubes.

On one hand, 58 Co surface activity at TRILLO 1 has decreased by a factor 3.8 on average inside primary system since cycle 4. On the other hand, 60 Co surface activity has increased by a factor 2.6 on average. These tendencies are consistent with those generally observed in French PWR [4.].

Dose rates

Dose rate measurements were performed around the RCS in order to estimate CEA indexes (figure 8) on primary piping and on SG tubes bundle. The results are reported in table 4. The average dose rate measured around primary system is higher in TRILLO 1 Cycle 21 than at St LAURENT B1 Cycle 20. The higher dose rate index of TRILLO primary piping is due to higher 60 Co surface activities (factor 2) and to a lower wall thickness (about 6 cm at TRILLO compared to about 7 cm at St-LAURENT). However, one can notice that the dose rate index at TRILLO would have been similar to St-LAURENT if the component thickness had been identical in spite of a higher level of 60 Co contamination because the ambient dose rate was lower at TRILLO than at St-LAURENT (contaminated by-pass pipes influence).

SG Tubes :

The lower dose rate index of the TRILLO SG tube bundle is due to a higher SG shell thickness (10 cm in TRILLO compared to 8.5 cm in St-LAURENT). The dose rate index of TRILLO would have been similar to the St-LAURENT one if the component thickness was identical.

CVCS Regenerative Heat Exchanger :

The dose rate close to the non regenerative heat exchanger is 10 times lower in TRILLO 1 than in St-LAURENT B1 : 0.23 mSv/h compared to 2.3 mSv/h in SLB120. This is essentially due to the contamination caused by 110m Ag which is the major contributor (about 65% to 80%) whereas 60 Co contribute for only 15% to 20%.

Contributors to dose rate :

As shown in figures 9 and 10, 60 Co is proved to be the major contributor to the dose rates around the primary system at CNT121 (between 80% and 90%), and SLB120 (between 65% and 75%). Contribution from 58 Co was about 5% in CNT121 and ranged between 10% and 25% in SLB120.

Around SGs, 124 Sb the contribution is not higher than 10%.

Collective dosimetry

The comparison between TRILLO 1 and St-LAURENT B1 indicates that TRILLO 1 had a very lower collective dosimetry than St-LAURENT B1 until 1999. This is due on one hand to higher contamination levels in 58 Co and 60 Co at St-LAURENT B1 mainly on SG tubes (see figure 2 to 5) and on the other hand probably to outage management. St-LAURENT B1 staff have performed a significant improvement program for the last decade, due to efficient ALARA plans. As a result of these actions, figure 9 shows quite similar collective dosimetry for both units in the last 3 years.

DISCUSSION

The 60 Co surface activity inside the primary system of TRILLO 1 at cycle 21 was higher than that of St-LAURENT B1 at cycle 20 and than that of Konvoy PWRs. The possible reasons for this higher 60 Co contamination level in TRILLO 1 are examined in this section hereafter.

First of all, a chemical conditioning such as the one at cycle 21 can not be the cause of this specific high 60 Co contamination level. In effect, it has been shown that primary chemistry may have a potential positive (and not negative) impact on contamination [5.] and there is no reason for the impact to only affect 60 Co.

Stellite hard-facing alloys could be responsible of this elevated 60 Co source term in case of abnormal wear, for instance during the grinding of valve seats. An abnormal wear of a component such as pump bearings (made of stainless steel with increased Co content) could also be a possible origin of contamination. However, an abnormal wear is generally characterized by the presence of hot spots inside the circuits, which is not the case at TRILLO 1.

Furthermore, Inconel 718 fuel grids could be the main cause of that observed 60 Co contamination. In fact, since the beginning, fuels assemblies at TRILLO Moreover, the nickel concentration has been high in operation since cycle 16 at least (between 1 and 10 ppb) and do not decrease as cycles go along. This level is not common, particularly in a PWR equipped with alloy 800 SG tubes (the usual values are measured between 0.1 and 1 ppb [9.]) and seems to show that Ni source term is "regenerated" after each outage, what could be consistent with refueling assemblies equipped with Inconel 718 grids.

Contamination levels at TRILLO-1 look, to a lesser extent, like those of BIBLIS-B, a pre-KONVOY PWR i.e. low 58 Co contamination levels (surface and released) and relatively high 60 Co surface activity for TRILLO-1 or very high for BIBLIS-B. The very high 60 Co contamination level in BIBLIS-B is due to the use of nickel-plated fuel spacers with an increased cobalt content [6.]. Inconel 718 fuel grids could also be the cause in TRILLO-1.

A priori, the SG tubing Co content is not responsible for this high 60 Co source term : the specified mean value was 150 ppm in TRILLO, which is similar to St-LAURENT-B1(measured content of 122 ppm) [4.].

Finally, The fact that the Ni concentration has remained at a high level for several cycles could be also due to the CVCS design : there is no particle filter upstream from the ion exchange resins. It seems that neither the mixed bed resins nor the mechanical filter downstream from them could trap particles efficiently.

CONCLUSION

The EMECC campaign performed during the beginning of the refuelling outage planned at the end of cycle 21 in TRILLO-1 revealed the following points in particular :

 The 58 Co surface activities on the primary coolant pipes and on the steam generator tubing in TRILLO-1 at cycle 21 were 2 times lower than those in St-LAURENT-B1 at cycle 20 and similar to those in the Konvoy PWRs. On the other hand, the Ni concentration has been abnormally high for several cycles in TRILLO-1.  The 58 Co peak during the outage 21 in TRILLO-1 was very low and similar to the one in SAINT-LAURENT-B1 at cycle 20 and to those in the Konvoy PWRs.  The reason for the low 58 Co surface activities and 58 Co peak in TRILLO-1 is the SG tubing alloy 800 (low Ni content), whereas the low 58 Co peak in St-LAURENT-B1 at cycle 20 is due to the alloy 690 SG tubing surface finish.  The 60 Co surface activities of the primary system were higher by a factor 2 in TRILLO-1 at cycle 21 compared to SAINT-LAURENT-B1 at cycle 20. The use of Inconel 718 fuel grids with a high Co impurity content must have played a great part in this high 60 Co contamination level.  The main contributor to the dose rates around the primary circuit in TRILLO-1 at cycle 21 was 60 Co, between 80% and 90%. As the 60 Co radioactive half-life is 5.3 years, the dose rates will remain relatively high for many cycles to come.

5.

Figure 1 .

 1 Figure 1. TRILLO 1 Cycle 1 -Cold shutdown procedure

Figure 2 .Figure 3 .Figure 4 .Figure 5 .

 2345 Figure 2. Comparison CNT121/SLB120 -58 Co and 60

Figure 8 .

 8 Figure 8. Primary piping and SG tube bundle : CEA dose index

Figure 9 . TRILLO 1 Figure 10 .

 9110 Figure 9. TRILLO 1 Cycle 21 -Radionuclide contribution to dose rates

Figure 9 :

 9 Figure 9 : Collective dose in TRILLO-1 and St-LAURENT B1 from 1989 to 2008

TRILLO -Unit 1 -Cycle 21 -Cold shutdown procedure

		350
		150 200 250 300			Zero power				Start of RHRS	Shutdown of RCPs						Start-up of RCS draining	3/4 Loop		RCS ventilation to atmosphere
	T (°C)	100
		50
		0	18	20	5	7	1	3	13 9		14 10	8	2	6	4	17	12	11	15	19	16	21
		10/02/09 12:00	11/02/09 00:00		11/02/09 12:00	12/02/09 00:00	12/02/09 12:00	13/02/09 00:00	13/02/09 12:00	14/02/09 00:00

-B (ppm/10) -Q (t/h) -H2 (ppm*100)

							2.8
							2.4
							2
							1.2 1.6	A (GBq/t) -O2 (ppm)
							0.8
							0.4
							0
	EMECC measurement	Temperature	Letdown flowrate	Boron	H2	O2	Co-58 volume activity

Co (1.33 GBq/t), the last RCP pump shutdown occurred only 2.5 hours after RHRS connection.

Table 1 . Comparison CNT121/SLB120 -Average volume activities during cycle

 1

	MBq/t	CNT 121	SLB 120
	58	Co	1.8	3.1
	60	Co	1.2	1.8
	54 Mn	0.9	1.5
	51	Cr	6.4	8.9

Table 2 . Average surface activities for some radionuclides in TRILLO 1

 2 Average surface activities measured at St LAURENT B1 cycle 20 are presented in table3for legs, SG tubes and CVCS letdown line.

	GBq/m²	58 Co	60 Co	54 Mn	124 Sb	110m Ag	58 Co/ 60 Co ratio
	Legs	0.76	2.69	0.16	0.08		0.28
	SG tubes	0.2	0.49		0.07		0.4
	CVCS letdown line (NRHE)	0.004	0.006		0.004	0.022	0.66

Table 3 . Average surface activities for some radionuclides in St LAURENT B1

 3

	GBq/m²	58 Co	60 Co	54 Mn	124 Sb	110m Ag	58 Co/ 60 Co ratio
	Legs	1.52	1.34	0.21	0.21	0.15	1.1
	SG tubes	0.53	0.29	0.04	0.07		1.8
	CVCS letdown line (NRHE)	0.04	0.04	0.006	0.047	0.32	1.0
	At TRILLO 1, one can observe that the main contributors to activity are 58 Co (≈ 20%) and 60 Co
	(≈ 65%) in RCS. The same radionuclides also mainly contribute to surface activity at St-
	LAURENT B1 but the 58 Co impact is significantly higher (40%-60% for 58 Co and ≈ 35% for
	60 Co) as shown by the different ratios 58 Co/ 60 Co.
	One can also notice a specific contribution of	110m Ag (≈ 50-60%) in CVCS non regenerative heat
	exchanger. A same tendency is observed at St-LAURENT B1. This phenomenon, well-known in
	French PWRs	[2.]

Table 4 . Comparison CNT121/SLB120 -CEA dose rate index

 4

	mSv/h	CNT 121 SLB120 CNT121 / SLB120
				Ratio
	Primary piping	0.49	0.27	1.8
	SG tube bundle	0.02	0.09	0.2
	CVCS NRHE	0.23	2.3	0.1
	Primary piping :

  1 have always contained Inconel 718 grids. During the four first cycles, all fuel spacer grids was made of Inconel 718. Afterwards and until cycle 17, only bottom and top grids were made of Inconel 718, the others grids being made in Zircaloy-4. More recently, the refueling assemblies have been made with a bottom fuel grid made of Inconel 718 and the others made of HPA-4. Since the Co specification of alloy 718 grids was high (1000 ppm), it is possible that the grids (activated by intense neutronic flux) have directly released Co into the primary coolant leading to a 60 Co contamination build-up.It should also be pointed out that60 Co and especially 58 Co surface activities were already relatively high at the end of cycle 4. Thus, this contamination, that probably occurred during the first cycles, could involve a nickel-based material as Inconel 718 grids (about 52% of Ni).

October 3-7, 2010


M Girard

F Dacquait

H Marteau

B Larat

F Nguyen

L Guinard

C Viala

The CIRENE loop: a tool to study ACP deposits and to validate the PACTOLE code

The understanding of Activated Corrosion Product (ACP) transfers in the primary circuit is of great importance for the PWR industry. For this purpose, the CEA, in collaboration with EDF and AREVA NP, has launched an R&D program for many years involving specific experiments related to the development and validation of the PACTOLE code. Among them, the CIRENE out-of-pile loop is dedicated to the study of ACP deposits on fuel rods and on steam generator tubes. The first results were presented at the Jeju Island International Conference. This paper presents the experimental results of a new CIRENE test and compares the numerical simulations obtained with the PACTOLE V3.2 code. This test has been carried out with thermal-hydraulic parameters corresponding to a subcooled boiling regime at the outlet of the core section, and chemical conditions representative of a French PWR primary circuit. A radioactive tracing methodology have been applied, using gamma spectrometry measurements with the injection of specific radiotracers, 58 Co and 59 Fe: in-situ gamma spectrometry measurements are performed at the outlet of the core section and the SG tubes, while frequent radiochemical analyses of the primary fluid are carried out for checking the instantaneous mass balances.

The comparison between the experimental results and simulations with the PACTOLE V3.2 code has led to the main following comments:

• deposited activity variations in operation : for the core section as well as for the SG tube section, the calculated variations reproduce the experimental ones;

• residual deposited activities after shutdown : the calculated activities are coherent with the experimental measurements. Consequently, the modelling developed in the PACTOLE code is globally validated for the CIRENE loop, and the differences between simulation and experimental results point out some remaining issues in the modelling of ACP behaviour in PWRs.

INTRODUCTION

The CIRENE loop is devoted to study the formation, migration and dissolution of corrosion product deposits on the component parts of a Pressurised Water Reactor, such as the fuel assemblies or the steam generator tubes.

The test was conducted to support the qualification of the PACTOLE code developed at CEA, the objective of the PACTOLE code (French acronym for Prediction of the behaviour and activation of corrosion products in the coolant system of light water reactors) being to predict the contamination in corrosion products in a PWR primary circuit.

This document presents a comparison between experimental results and numerical simulations of this test with the version V3.2 of the PACTOLE code [START_REF] Dacquait | Simulations of corrosion product transfer with the PACTOLE V3.2 code[END_REF], the test conditions and the CIRENE loop modelling are specified.

THE CIRENE LOOP

The CIRENE loop [START_REF] Yagnik | CRUD deposition studies in the CIRENE loop[END_REF] is dedicated to study the formation of corrosion product deposits on a "core" section consisting of 4 fuel claddings and on a "SG tube" section consisting of 4 steam generator tubes. This pressurised water circulation loop corresponds to a PWR coolant system at a scale of 1/7000. The thermal hydraulic and chemical environment of the PWRs can be reproduced there, with transfer phases corresponding to various operating conditions. The presence of fuel is simulated by heater components inserted in each cladding. An auxiliary heater is used to reach the required inlet core temperature. A thermal pressurizer and a primary circulation pump are used to maintain respectively the fluid pressure and the fluid velocity in the RCS. The outlines of the CIRENE facility required for the test are presented on Figure 1.

M. Girard, F. Dacquait, H. Marteau, B. Larat, F. Nguyen, L. Guinard, C. Viala, "The CIRENE loop : a tool to study ACP deposits and to validate the PACTOLE code", Proceedings of the International Conference on Water Chemistry of Nuclear Reactors Systems, 15-18/09/08, Berlin (Germany), P1-05, 2008

The "core" section [ SE1, SE2, SE3, SE4 ]

The "core" section is composed of four Zircaloy-4 alloy claddings, in contact with the primary coolant over a length of 3.40 m. Seven Zircaloy-4 grids distributed axially support the claddings inside a stainless steel 316L tube of square cross-section. The geometric characteristics of the grids are those of a 17x17 PWR assembly. The average length of each cladding subjected to the heat flux is 1.80 m.

The heat exchanger "SG tube" section [ ECC, ECT, ECF ]

The Steam generator section is simulated by the primary/secondary heat exchanger of the CIRENE loop constituted by four industrially manufactured Inconel-690 TT tubes (non-electropolished and thermallytreated). This is a single-pass straight tube heat exchanger.

Corrosion product injection device [BINJ_ Fe, BINJ_ Ni]

This injection device is used to inject iron (as Fe 2+ ) and nickel (as Ni 2+ ) solutions in the primary coolant so as to be representative of the potential contamination of a French PWR. These solutions are continuously injected into the primary circuit via the CVCS (chemical and volume control system) at the inlet of the "core" section. This device is dimensioned to simulate total iron and nickel release rates of between 0.6 to 2.5 10 -10 kg/m 2 .s (i.e 1.6 to 6.4 mg/dm 2 .month).

Radioactive tracer injection device [BINJ_ Fe59, BINJ_Co58]

This device is used to performed radioactive tracer injections into the primary coolant in soluble form. The radiotracers selected for this test, respectively 58 Co and 59 Fe, have been chosen for their adapted half-life periods and gamma energies while being part of the most relevant Activated Corrosion Products (ACP) in PWR. During the present test, series of radiotracer injections are simultaneously performed while continuously injecting the inactive precursors.

THE TEST OPERATION

The thermal hydraulic conditions correspond to an under-saturated boiling regime at the upper part of the "core" section and are kept constant during the test. The chemical conditioning of the primary fluid is representative of PWR operation, a variation of the pH number and of the hydrogen content is carried out during the test. The formation and migration of corrosion products are monitored by injecting radioactive tracers combined with γ spectrometry measurements on two different areas of the RCS and using samples of the primary fluid.

Thermal hydraulic and chemical conditions

The thermal hydraulic parameters maintain at the outlet of the "core" section an evaporation rate of about 0.36 kg/m 2 /s (i.e 1 300 kg/m 2 /h) in the average sub-channel. The thermal hydraulic data are given in Table I below. The vaporization rate and the void fraction are estimated using the 3D thermal-hydraulics code FLICA IV [START_REF] Gallo | FLICA 4 -Version 1 -Manuel de référence de la modélisation[END_REF] developed by CEA. concentrations can be monitored using probes on the CVCS, as well as the pH number and the electrochemical potential. Three specific injections of soluble nickel are undertaken in step 2 with an injection rate 50 times higher than the representative release rates mentioned above: the aim is to reach the saturation level of this element at the inlet of the core section and therefore to precipitate along the "core" section. The Table 3 below gives the CP injection data.

gamma spectrometry measurements and analyses of the deposit

Gamma spectrometry measurements implemented for this test concern

• in-situ and in-line measurements on RCS areas;

• primary fluid samples measurements;

• after shutdown measurements.

As specified on Figure 1, the in-line measurement areas are performed at the outlet of both the core section and the SG tubes, using two High Purity Ge semi-conductor detectors: a coaxial 100% efficiency probe located at the upper part of the core section, and a coaxial 40% efficiency probe located at the outlet of the SG tubes.

For each measured spectrum, the 58 Co peak at 811 keV and the 59 Fe peaks at 1099 and 1292 keV are considered. The counting periods have been fixed at 3 600 s. The methodology applied for γ measurements during this test is detailed in the paper presented at the Jeju Island International Conference [START_REF] Blet | In-situ ACP deposit measurements in the CIRENE out-of-pileloop[END_REF]. The frequent measurements of the primary coolant activity are carried out with γ spectrometry of 50 ml samples using a 20% efficiency HP Ge detector. Additional γ measurements are undertaken after shutdown and dismantling: measurements on Zr-4 claddings, on the 316L square tube and on CRUD scrapings at various axial elevations. Activity of the remained deposit (CRUD), as well as the activity of the labile remained deposit, is therefore obtained. Scrapings are also subjected to chemical analyses (Xray fluorescent spectrometry) in order to determine the composition in CP elements in the labile remained deposit. The relevant surface density can be estimated by weighing.

MODELLING OF THE CIRENE LOOP FOR THE PACTOLE CODE V3.2

Schematization of the loop

The schematization of the CIRENE loop required for PACTOLE simulations is presented on Figure 2. The loop is represented by 35 control volumes. The total volume of this circuit is 2.8.10 -2 m 3 . The total wet surface, excluding injection devices [ BINJ ] is 3.87 m². The parameters Wet area, HDiam., Tfluid and Twall represent respectively the wet surface (in m²), the hydraulic diameter (in m), the average fluid temperature (en °C) and the wall temperature (in °C). The meaning of the abbreviations is listed below:

• SE: "core" section divided into 22 control volumes Injections in the primary coolant occur at a point located between [Aux. heater] and [TU4], upstream from the "core" section.

Input data set

The basic data set is compatible with version V3.2 of the code, the parameters of the physical models are those used for French PWR plants simulations.

• "Core" section This section consists of four Zircaloy-4 claddings and a stainless steel 316L square tube: a material modelling the mixture of the two must therefore be created, respecting the proportions in terms of wet surfaces. The material INZR was therefore defined for the "core" section, with 45.6 % stainless steel and 54.4% Zircaloy in wet surfaces.

The temperatures in the primary coolant and on the cladding walls are determined from a FLICA IV calculation [START_REF] Gallo | FLICA 4 -Version 1 -Manuel de référence de la modélisation[END_REF] for the 22 control volumes from [SE1_a] to [SE4_j]. Figure 3 describes the temperature profiles integrated in the input data set for this section with the relevant control volumes.

• ACP and CP injections These injections are expressed in terms of release via data expressed in g/s. These volumes are subjected to neither deposit nor corrosion, and release only iron, nickel, 59 Fe and 58 Co.

• Complete description of the evolution of the primary coolant conditioning in B, Li and H 2 during the 30.5 days of the test.

COMPARISON BETWEEN EXPERIMENTAL MEASUREMENTS AND PACTOLE V3.2 SIMULATIONS

Results of the CIRENE test simulations

The calculation stops at t = 30.5 days operation. It does not include any shutdown phase. The simulation lasted 12 min for a 4h-time step calculation on a computer with the following configuration: Intel Pentium 4 -CPU 3.2 GHz -1 GB RAM -Windows OS.

Deposited activity profiles

The profiles of the in-line γ spectrometry measurements are compared with the calculation results for the [SE4_i] "core" outlet and for the [ ECF ]: SG tube outlet control volumes on Figures 4, 5 and6. The tracer injection sequences are displayed in terms of total activity injected in the RCS. Note that the in-line measurements on the "core" section are given in counts per second and that they cannot be directly compared with the surface activity values given by the simulation. This limitation is due to the fact that, a priori, the distribution of activities in the core section between the claddings, the square tube and the force tube is unknown. The in-line measurements on the SG tube section can be compared directly with the surface activity values of the simulation, they are given in Bq/cm 2 (1 Bq/cm 2 = 1E-05 GBq/m 2 ).

For both the core section and the SG tube section, the "Calculation" profiles exhibit good agreement with the "Measurement" profiles. The "Calculation" profiles effectively exhibit the 11 radioactive tracer injection sequences. The increase gradients calculated and measured during the injections are similar. Between 2 injection phases, however, the deposited activities continue to increase on the core section, but with a lower gradient, whereas the PACTOLE calculation indicates stability or even a slight decrease due to radioactivity decay. In contrast, concerning the SG tubes, as soon as injection stops the deposited activities stop increasing, or even decrease, which is more in agreement with the PACTOLE calculation.

The calculated levels of the deposited activities in 58 Co and in 59 Fe at t = 30.5 days in the SG tubes are very close to the measurements, by a factor of less than 2. However for the Co-58, the code tends to underestimate the deposited activities on the SG tubes during the injection phase.

On the 18 th day of the test, as the reduction in hydrogen and boron occur (start of step 3), a very slight increase in the core section contamination gradient can be noticed which PACTOLE fails to represent. Deposited activity at 30.5 days Assuming that the rapid shutdown with ∆T=197°C/10 min without any oxidation led to a weak dissolution of the deposits in the upper part of the core section, then for this area, calculated activities at 30.45 days could be compared with those measured after shutdown. The values to be compared are shown in Table 5 below. We observe a good agreement between the "Calculation" and "Measurement" values, and more especially on the SG tube section, which had already

The calculation stops at t = 30.45 operation days. It includes a complete description of the evolution of the primary coolant conditioning in B, Li and H 2 , as well as the sequences corresponding to the ACP (Co-58 and Fe-59) and CP (Fe and Ni) injections.

The CIRENE test calculations give the following results:

Deposited activity profiles: note that, for both the core section and the SG tube section, the "Calculation" profiles display good agreement with the "Measurement" profiles. The "Calculation" profiles effectively retranscribe the 11 radioactive tracer injection sequences, although sometimes the kinetics and amplitudes are different. Deposited activity at 30.5 days: we observe good agreement between the "Calculation" and "Measurement" values, especially on the SG tube section. Deposited masses at 30.5 days: we obtain the same order of magnitude between "Calculation" and "Measurement" (i.e. the surface density of the labile remained deposit) in the upper part of the "core" section, whereas we observe a difference as soon as we examine the control volumes at lower axial elevations of the under-flux area. Composition of the primary coolant in ACP and CP elements : for the non-radioactive as well as for the radioactive elements, the experimental values and the calculation agree, but only to a certain extent since the detection limits of the required analysis methods are often reached. It is important to note that γ spectrometry measurements combined with injection of radiotracers provide access to numerous experimental data and profiles which give added value in the comparison with code simulations. This CIRENE test and the relevant simulations have revealed new directions in the understanding of contamination transfer phenomena and ways of improving their modelling in the PACTOLE code (contamination kinetics, temperature effects, chemical conditioning effects…).

SE 4 _ j W et area 1 .6 8 E -0 1 W et area 2 .3 E -0 1 H D iam . 2 .4 3 E -0 2 W et area 2 .4 2 E -0 1 W et area 2 .4 2 E -0 1 W et area 2 .4 2 E -0 1 H D iam . 6 .2 2 E -0 3 T flu id 3 3 1 H D iam . 1 .7 0 E -0 2 H D iam . 1 .7 0 E -0 2 H D iam . 1 .7 0 E -

Figure 1 :

 1 Figure 1: CIRENE loop outlines required for the test The primary circuit components involved in the modelling by the PACTOLE code are listed below, the relevant regions or control volumes being commented [between brackets]:

•

  ECX: heat exchanger section divided in 3 control volumes where C = hot part, T = middle part and F = cold part • TU: tube connecting the various parts of the loop • DER: bypass of a radioactive area (active knitmeshes used during previous tests ) • Aux. heater : the auxiliary heater • BINJ (Wet area = 1 m 2 and H Diam.= 2.43.10 -2 m): this region includes 4 control volumes [BINJ_Fe, BINJ_Ni, BINJ_Fe59 and BINJ_Co58] corresponding to the ACP and CP injection devices.

Figure 2 :Figure 3 :

 23 Figure 2: CIRENE loop schematization for PACTOLE code simulations

Figure 7 : 3 Figure 8 :Figure 9 :

 7389 Figure 7: Primary fluid activities in Fe-59 and Co-58 calculated for the CIRENE test ([SE_1a] control volume -PACTOLE V3.2 code / 4h-time step calculation)

Table 1 : thermal hydraulic parameters of the CIRENE test

 1

	Pressure	15 MPa
	Primary flow rate	1.3 10 -3 m 3 /s
	Mass flow rate in the core section	2 350 kg/m 2 /s
	Core section -inlet temperature	300.5 °C
	-outlet temperature	330.5 °C
	Thermal flux	75.10 -4 W/m 2
	FLICA IV calculations at the outlet of
	the core section (average sub-channel)
	Outlet temperature Rate of vaporization	331 °C 0.36 kg/m 2 /s
	Void fraction	~ 0.6 %
	The nominal chemical conditions are summarized in
	Table 2 below. The pH and the hydrogen content of the
	primary coolant set to 6.8 and 30 cm 3 /kg at the
	beginning of the test, are then modified with a pH
	increase scheduled in step 3 and an increase in the H 2
	content up to 55 cm 3 /kg in step 4. Oxygen and hydrogen

Sampling Primary pump Auxiliary heater RCS Pressurizer (24 L ) 4 x claddings (Zry4) and 316L SSteel square tube CP injection device (Fe, Ni) Primary heat exchanger 4 x SG tubes ( Inconel 690TT) T H 2 O Li, B Make up tank (200 L) secondary circuit ACP injection device (Fe-59,Co-58)

				Make up Make up
				tank (200 L) tank (200 L)
	secondary
	circuit
			Pressurizer	H 2 O H 2 O
		Primary heat exchanger	(24 L )	Li, B Li, B
	T	4 x SG tubes ( Inconel 690TT)
	in-line γ γ γ γ spectrometry areas in-line γ γ γ γ spectrometry areas	T T
	4 x claddings (Zry4) and 316L SSteel square tube	RCS
	Auxiliary heater
	T T	Primary pump	CP injection device
	T T			(Fe, Ni)
	Sampling	CVCS CVCS		ACP injection
				device
				(Fe-59,Co-58)

Table 2 :

 2 chemical conditioning of the primary fluid during the CIRENE test

		Step 0 to step 2 Step 3	Step 4
	Duration	18 days	8 days	5 days
	B /Li (mg/kg)	2 100 / 1.3	140 / 0.8 140 / 0.8
	pH (300°C)	6.8	7.2	7.2
	O 2 (µg/kg)	< 10	< 10	< 10
	H 2 (cm 3 /kg)	24 -28	30 -40	40 -55
	3.2 External injections of Corrosion Products and
	Activated Corrosion Products
	Injection of iron and nickel in soluble states (CP)

The solutions are sequentially injected throughout the test, with a Ni/Fe ratio maintained between 2.5 and 4.3, the injection rate being representative of a nickel alloy release in PWR environment.

Table 3 :

 3 specific setting for the CP injection device The 58 Co and 59 Fe solution was injected in 11 sequences throughout the test with the specific setting details in Table4. The average injection rate of the radioactive tracer solution, in terms of injected activity is 0.01 kBq/s. By the end of the 30 days test, 1 940 kBq in58 Co and 59 Fe had been injected for a total mass of 1.36 ng.

	Radiotracer	Number	Injection rate	Total
		of sequences	in mg/s	injected masses
	Fe	16	3.05E-04	270 mg
	Ni	15	10.3E-04	730 mg
	Ni	3	500E-04	120 mg
	Injection of radioactive tracers in soluble states (ACP)
	The radiotracer injections are simultaneously performed
	while continuously injecting the inactive CP, each ACP
	injection lasting ~ 5 hours.
	Table 4: specific setting for the ACP injection device
	Radiotracer	Number	Injection rate	Total
		of sequences	in kBq/s	injected masses
	Fe-59	11	7E-03	0,36 ng
	Co-58	11	3.5E-03	1 ng

Table 5 :

 5 Deposited activities calculated at t=30.5 days and measured after rapid shutdown for the CIRENE test, activities are given in Bq/cm 2 i.e. 1E-05GBq/m 2 .

			Experimental
		PACTOLE V3.2	measurements
		calculation	after shutdown on
			claddings + square tube
		Control volume [ SE4_i] : core outlet
	Co-58	75	82 +/-27%
	Fe-59	40	50 +/-15%
		Control volume [ ECF ] : SG tubes outlet
	Co-58	13	12 +/-3%
	Fe-59	1.7	1.5 +/-9%

ACKNOWLEDGEMENTS

The authors are thankful to the CIRENE loop staff V. BLET, G. FRITZ, A. BLANC and to N. GALLIANO, A. ALI, S. BANTICHE who performed the gamma measurements.

been demonstrated by direct comparison of the activity profiles (see figure 6). Besides an overall view of the contamination levels calculated at t = 30.5 days in the 35 control volumes of the CIRENE loop points out that the core section is the most contaminated area and that the contamination level of the out-of flux area [SE1_a] to [SE1_d] is quite similar to that of the under-flux area [SE_2] to [SE_4].

Deposited masses on the "core" section We compare in Table 6 the masses of deposited corrosion products calculated at 30.5 operation days with those measured after shutdown using characterizations of the cladding and square tube scrapings. Note that this relates to the characterization of a labile remained deposit resulting from a rapid shutdown without oxygenation. The relevant control volumes are [SE2], [SE3] and [SE4], corresponding to the under-flux area as depicted on Figure 3. We obtain the same order of magnitude of surface density in the upper part of the claddings, whereas a difference can be observed as soon as we examine the control volumes corresponding to the beginning of the under-flux area: it can be reasonably assumed that the rapid shutdown has led to a partial CRUD dissolution in the lower part of the core section where oxides of ferrite nickel type are preferentially formed, whereas it has preserved the Ni-enriched CRUD deposited in the upper part of the core section [START_REF] Yagnik | CRUD deposition studies in the CIRENE loop[END_REF].

Composition in soluble CPs and ACPs in the primary coolant The experimental results are compared with the calculated concentrations and with equilibrium concentrations C eq obtained using the PHREEQC-CEA code integrated in the PACTOLE code V.3.2 (see Figures 7 and8). Experimental analyses conducted in step 1 and 2 give the results shown in Table 7 below.

For the non-radioactive CPs as well as for the ACPs, the experimental values and the calculation agree, but only to a certain extent since the detection limits of the analysis methods implemented are often reached (Chemical analysis DL: from 2 to 2.5.E-06 kg/m 3 and Radiochemical analysis DL : from 4.1E+03 to 7.5E+03 Bq/m 3 ).

Mechanisms causing the increase in deposited activity

To determine the main mechanisms of PACTOLE V3.2 causing the increase in deposited activities on the claddings and the SG tubes, calculations were carried out successively deleting the precipitation mechanisms on the "particle" environment and on the "deposit" environment.

We observe that deposited activities on the claddings are due to the phenomenon of radioactive particle deposition as can be observed in Figure 9. The radioactive ions injected precipitate on the particles, which are then deposited on the claddings. As can be seen in Figure 10, the same behaviour is observed for the 59 Fe deposited activity profile in the SG tube section. In contrast, the 58 Co deposited activity is mostly due to precipitation on the surface of the deposit in the SG tube section.

CONCLUSIONS

This paper presents the simulation results of the CIRENE test in the version V3.2 of the PACTOLE code whose are compared with the available experimental measurements. Note that this test was carried out with thermal hydraulic conditions at subcooled boiling limit at the outlet of the "core" section, and that nonradioactive and radioactive corrosion products were injected in the primary coolant.