# Simulations of corrosion product transfer with the PACTOLE V3.2 code

Frédéric Dacquait, Frédéric Nguyen, Hervé Marteau, Bertrand Larat, Dominique You, Gabriel Plancque, Luc Guinard, Gilles Ranchoux, Céline Viala, Martin Bachet

► To cite this version:

Frédéric Dacquait, Frédéric Nguyen, Hervé Marteau, Bertrand Larat, Dominique You, et al. Simulations of corrosion product transfer with the PACTOLE V3.2 code. NPC 2008 - International Conference on Water Chemistry of Nuclear Reactors Systems, Sep 2008, Berlin, Germany. cea-04474054

HAL Id: cea-04474054

https://cea.hal.science/cea-04474054

Submitted on 22 Feb 2024

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Distributed under a Creative Commons Attribution 4.0 International License
# Simulations of corrosion product transfer with the PACTOLE V3.2 code

Frédéric Dacquait, Frédéric Nguyen, Hervé Marteau, Bertrand Larat, Dominique You, Gabriel Plancque, Luc Guinard, Gilles Ranchoux, Céline Viala, Martin Bachet

► To cite this version:

Frédéric Dacquait, Frédéric Nguyen, Hervé Marteau, Bertrand Larat, Dominique You, et al. Simulations of corrosion product transfer with the PACTOLE V3.2 code. NPC 2008 - International Conference on Water Chemistry of Nuclear Reactors Systems, Sep 2008, Berlin, Germany. cea-04474054

HAL Id: cea-04474054

https://cea.hal.science/cea-04474054v1

Submitted on 22 Feb 2024

HAL is a multi-disciplinary open access
archive for the deposit and dissemination of sci-
entific research documents, whether they are pub-
lished or not. The documents may come from
teaching and research institutions in France or
abroad, or from public or private research centers.

L'archive ouverte pluridisciplinaire HAL, est
destinée au dépôt et 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.

<figure><img src="image_1.png" /></figure>

Distributed under a Creative Commons Attribution 4.0 International License
# Simulations of corrosion product transfer with the PACTOLE V3.2 code

F. Dacquait¹, F. Nguyen¹, H. Marteau¹, B. Larat¹*, D. You¹, G. Plancque¹, L. Guinard², G. Ranchoux², C. Viala³, M. Bachet⁴

¹CEA/DEN, ²EDF/SEPTEN, ³AREVA NP, ⁴EDF/R&D

CEA, DEN, F-13108 Saint-Paul-Lez-Durance, France.

E-mail: frederic.dacquait@cea.fr

## Abstract

Activated corrosion products generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfers is of the highest importance. For about thirty years, the French strategy has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development.

The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection and activation. In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300°C, 150 bar, neutron flux, water velocity up to 15 m.s⁻¹ and very low corrosion product concentrations).

The purpose of the PACTOLE code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the contamination of the PWR primary system. The PACTOLE code allows researchers to analyse the corrosion product behaviour and to calculate the activity in the fluid and the surface activity in the primary system. Nowadays, the PACTOLE code is considered to be not only a tool for numerical simulations and predictions but also one that might combine and organize all new knowledge useful to progress on contamination caused by activated corrosion products.

This paper presents the modelling implemented in the PACTOLE V3.2 code. In this version, the chemical aspect has been improved by coupling the PACTOLE code to a chemistry code. Comparisons with in-situ gamma spectrometry measurements of PWR primary systems and studies based on experimental feedback are shown.

## 1. INTRODUCTION

Deposits of activated corrosion products found in the primary system of Pressurized Water Reactors (PWR) generate a radiation field responsible for the doses absorbed by personnel carrying out maintenance and reactor operation. It is therefore very important to understand the mechanisms governing the transfer of corrosion products. For about thirty years now, the French strategy has been focusing on 1) performing experiments in test loops that recreate PWR conditions, 2) taking in situ measurements of deposits in PWR systems, and 3) developing a simulation code known as PACTOLE.

The PACTOLE code has been developed by the CEA in collaboration with EDF and AREVA NP for more than 30 years [1], [2], [3], [4], [5]. It is designed to simulate the transfer of corrosion products in the primary system of a PWR. It determines the space-time distribution of masses and/or activity levels of corrosion products and activation products in both solid and liquid phases in the primary system.

This paper aims to describe the modelling used to simulate the transport of corrosion products as

implemented in the PACTOLE V3.2 code. It also provides comparisons between simulations and experimental measurements performed in PWRs.

## 2. PWR CONTAMINATION PRINCIPLE

The process governing the formation of radioactive corrosion products in the PWR primary system is complex and involves several interacting mechanisms.

The first mechanism is the uniform and generalized corrosion of the base metal composing components in contact with the primary coolant, i.e. mainly stainless steels and alloys containing nickel or iron. This results in the formation of double-layered oxide: a compact inner layer that is rich in chromium (chromite) and an outer porous layer that is rich in iron (ferrite). The inner layer is passive, which limits the exchange of ions between the base metal and primary coolant but without preventing them completely; some ions are released in the primary coolant.

The primary coolant transports the ions produced by corrosion and release phenomena or by the dissolution of oxides. When the coolant becomes oversaturated in corrosion products (depending on the local temperature and chemical conditions), the ions either precipitate on the surface of the walls, or precipitate to form particles in

<footnote>* Present affiliation : Euriware</footnote>
# Simulations of corrosion product transfer with the PACTOLE V3.2 code

F. Dacquait¹, F. Nguyen¹, H. Marteau¹, B. Larat¹*, D. You¹, G. Plancque¹, L. Guinard², G. Ranchoux², C. Viala³, M. Bachet⁴

¹CEA/DEN, ²EDF/SEPTEN, ³AREVA NP, ⁴EDF/R&D

CEA, DEN, F-13108 Saint-Paul-Lez-Durance, France.

E-mail: frederic.dacquait@cea.fr

## Abstract

Activated corrosion products generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfers is of the highest importance. For about thirty years, the French strategy has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development.

The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection and activation. In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300°C, 150 bar, neutron flux, water velocity up to 15 m.s⁻¹ and very low corrosion product concentrations).

The purpose of the PACTOLE code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the contamination of the PWR primary system. The PACTOLE code allows researchers to analyse the corrosion product behaviour and to calculate the activity in the fluid and the surface activity in the primary system. Nowadays, the PACTOLE code is considered to be not only a tool for numerical simulations and predictions but also one that might combine and organize all new knowledge useful to progress on contamination caused by activated corrosion products.

This paper presents the modelling implemented in the PACTOLE V3.2 code. In this version, the chemical aspect has been improved by coupling the PACTOLE code to a chemistry code. Comparisons with in-situ gamma spectrometry measurements of PWR primary systems and studies based on experimental feedback are shown.

## 1. INTRODUCTION

Deposits of activated corrosion products found in the primary system of Pressurized Water Reactors (PWR) generate a radiation field responsible for the doses absorbed by personnel carrying out maintenance and reactor operation. It is therefore very important to understand the mechanisms governing the transfer of corrosion products. For about thirty years now, the French strategy has been focusing on 1) performing experiments in test loops that recreate PWR conditions, 2) taking in situ measurements of deposits in PWR systems, and 3) developing a simulation code known as PACTOLE.

The PACTOLE code has been developed by the CEA in collaboration with EDF and AREVA NP for more than 30 years [1], [2], [3], [4], [5]. It is designed to simulate the transfer of corrosion products in the primary system of a PWR. It determines the space-time distribution of masses and/or activity levels of corrosion products and activation products in both solid and liquid phases in the primary system.

This paper aims to describe the modelling used to simulate the transport of corrosion products as

implemented in the PACTOLE V3.2 code. It also provides comparisons between simulations and experimental measurements performed in PWRs.

## 2. PWR CONTAMINATION PRINCIPLE

The process governing the formation of radioactive corrosion products in the PWR primary system is complex and involves several interacting mechanisms.

The first mechanism is the uniform and generalized corrosion of the base metal composing components in contact with the primary coolant, i.e. mainly stainless steels and alloys containing nickel or iron. This results in the formation of double-layered oxide: a compact inner layer that is rich in chromium (chromite) and an outer porous layer that is rich in iron (ferrite). The inner layer is passive, which limits the exchange of ions between the base metal and primary coolant but without preventing them completely; some ions are released in the primary coolant.

The primary coolant transports the ions produced by corrosion and release phenomena or by the dissolution of oxides. When the coolant becomes oversaturated in corrosion products (depending on the local temperature and chemical conditions), the ions either precipitate on the surface of the walls, or precipitate to form particles in

<footnote>* Present affiliation : Euriware</footnote>
the coolant. Particles are also produced by the erosion of deposits. When transported by the primary coolant, these particles can agglomerate and deposit on the primary system walls.

Activation products can be formed in two different ways: 1) by activation in the core when the corrosion products form a deposit on surfaces or circulate in the system under the neutron flux, or 2) via the corrosion of components under the neutron flux, including the direct release of radioactive corrosion products.

These activated corrosion products can then form a deposit on the out-of-flux surfaces and create an ionising radiation field around PWR systems.

### 3. MODELLING OF THE PACTOLE V3.2 CODE

#### 3.1. Representation of the primary system

PACTOLE V3.2 modelling is based on the arrangement of control volumes (see Figure 1).

Therefore, the PWR primary system is discretised into several volumes or regions defined according to the geometric, thermohydraulic, neutron, material property, and operating characteristics.

The primary system is composed of a main line (blue volumes in Figure 1) to which various lines and circuits are connected, e.g. the chemical and volume control system (CVCS). The loops of a PWR primary system are generally considered as one main loop (see Figure 1) but it is possible to differentiate between them to take into account, for example, the state of a different surface of the steam generator tubes.

<figure><img src="image_2.png" /><figcaption>Figure 1: Representation of a PWR primary system and a CVCS in terms of control volumes</figcaption></figure>

Seven media can be taken into account in each control volume (see Figure 2): the base metal, the inner oxide layer, the outer oxide layer, the deposit, the ions and particles carried by the coolant, and the filters for particles and ions.

<figure><img src="image_3.png" /><figcaption>Figure 2: Media taken into account in PACTOLE code modelling</figcaption></figure>

**Base metals:** alloys composing the different elements of the primary system. The metal in contact with the primary coolant is subjected to the corrosion process.

**Inner oxide layer:** described as a rather adhesive layer composed of small-sized grains. This oxide coats all surfaces in contact with the primary coolant. Its chemical composition is rich in chromium. When coating stainless steels and nickel-based alloys, this oxide is a chromite according to the following formula: $Co_xNi_yFe_{1-x-y}Cr_2O_4$ with $x+y\le 1$.

**Outer oxide layer:** formed by crystal agglomerates produced by the surface precipitation of ions in the primary coolant [6]. It is assumed that the surface of the inner oxide layer is coated by these crystals. In the case of nickel- and iron-based alloys for example, this oxide is a mixed oxide whose formula can be expressed as follows: $Ni_xCo_yFe_{3-x-y}O_4$ with $0\le x\le 1$ and $0\le y\le 1-x$. In relation to the values of x and y, this can be magnetite, nickel ferrite, or cobalt ferrite. Metallic nickel (Ni) or nickel oxide (NiO) may also be found.

**Deposit:** deposits are composed of oxide particles resulting from 1) the settling of particles carried by the primary coolant, and 2) the precipitation (on the surface of the previously-formed deposit) of ions in the coolant. The deposit is modelled as a porous medium made of spherical particles whose particle size distribution is identical to those of the particles suspended in the coolant. The deposit is superimposed on the outer oxide layer and can partially or completely coat this layer.

**Particles:** carried by the primary coolant and resulting from the precipitation of ions in the coolant. They can also be produced by the erosion of deposits and the outer oxide layer. We assume that the particles are spherical and that their grain size distribution follows the "log-normal" law.

**Ions:** the primary water contains ions that are transported by convection in the primary system. The ions in the primary coolant result from corrosion release phenomena in the bulk of the coolant, and due to the dissolution of particles, the outer oxide layer and deposits.

**Filters and resins:** these two systems are used to purify the primary system to remove the particles and ions.
the coolant. Particles are also produced by the erosion of deposits. When transported by the primary coolant, these particles can agglomerate and deposit on the primary system walls.

Activation products can be formed in two different ways: 1) by activation in the core when the corrosion products form a deposit on surfaces or circulate in the system under the neutron flux, or 2) via the corrosion of components under the neutron flux, including the direct release of radioactive corrosion products.

These activated corrosion products can then form a deposit on the out-of-flux surfaces and create an ionising radiation field around PWR systems.

### 3. MODELLING OF THE PACTOLE V3.2 CODE

#### 3.1. Representation of the primary system

PACTOLE V3.2 modelling is based on the arrangement of control volumes (see Figure 1).

Therefore, the PWR primary system is discretised into several volumes or regions defined according to the geometric, thermohydraulic, neutron, material property, and operating characteristics.

The primary system is composed of a main line (blue volumes in Figure 1) to which various lines and circuits are connected, e.g. the chemical and volume control system (CVCS). The loops of a PWR primary system are generally considered as one main loop (see Figure 1) but it is possible to differentiate between them to take into account, for example, the state of a different surface of the steam generator tubes.

<figure><img src="image_2.png" /><figcaption>Figure 1: Representation of a PWR primary system and a CVCS in terms of control volumes</figcaption></figure>

Seven media can be taken into account in each control volume (see Figure 2): the base metal, the inner oxide layer, the outer oxide layer, the deposit, the ions and particles carried by the coolant, and the filters for particles and ions.

<figure><img src="image_3.png" /><figcaption>Figure 2: Media taken into account in PACTOLE code modelling</figcaption></figure>

**Base metals:** alloys composing the different elements of the primary system. The metal in contact with the primary coolant is subjected to the corrosion process.

**Inner oxide layer:** described as a rather adhesive layer composed of small-sized grains. This oxide coats all surfaces in contact with the primary coolant. Its chemical composition is rich in chromium. When coating stainless steels and nickel-based alloys, this oxide is a chromite according to the following formula: $Co_xNi_yFe_{1-x-y}Cr_2O_4$ with $x+y\le 1$.

**Outer oxide layer:** formed by crystal agglomerates produced by the surface precipitation of ions in the primary coolant [6]. It is assumed that the surface of the inner oxide layer is coated by these crystals. In the case of nickel- and iron-based alloys for example, this oxide is a mixed oxide whose formula can be expressed as follows: $Ni_xCo_yFe_{3-x-y}O_4$ with $0\le x\le 1$ and $0\le y\le 1-x$. In relation to the values of x and y, this can be magnetite, nickel ferrite, or cobalt ferrite. Metallic nickel (Ni) or nickel oxide (NiO) may also be found.

**Deposit:** deposits are composed of oxide particles resulting from 1) the settling of particles carried by the primary coolant, and 2) the precipitation (on the surface of the previously-formed deposit) of ions in the coolant. The deposit is modelled as a porous medium made of spherical particles whose particle size distribution is identical to those of the particles suspended in the coolant. The deposit is superimposed on the outer oxide layer and can partially or completely coat this layer.

**Particles:** carried by the primary coolant and resulting from the precipitation of ions in the coolant. They can also be produced by the erosion of deposits and the outer oxide layer. We assume that the particles are spherical and that their grain size distribution follows the “log-normal” law.

**Ions:** the primary water contains ions that are transported by convection in the primary system. The ions in the primary coolant result from corrosion release phenomena in the bulk of the coolant, and due to the dissolution of particles, the outer oxide layer and deposits.

**Filters and resins:** these two systems are used to purify the primary system to remove the particles and ions.
## 3.2. Chemical elements and isotopes

The chemical elements taken into account in the PACTOLE V3.2 code are those composing the main alloys found in any primary system that generate the main activated corrosion products: Fe, Ni, Cr and Co.

Certain isotopes of these chemical elements become radioactive when subjected to a neutron flux. For example, the main activated corrosion products that contribute to the dose rates around the primary systems in EDF PWRs are ⁵⁸Co and ⁶⁰Co with half-lives of 70.78 days and 5.271 years respectively [7]. They are formed by the following reactions:

* ⁵⁸Ni (n,p) ⁵⁸Co

* ⁵⁹Co (n, γ) ⁶⁰Co

The mass balances are performed for each isotope of these chemical elements.

## 3.3 Mass balance and transfer mechanisms

The behaviour of the corrosion products in each control volume is modelled by the mass balance equation for each isotope in each medium [4]:

$$ \frac{\partial m_i}{\partial t} + (\dot{m}_{out} - \dot{m}_{in}) = \sum_{Source} J_m - \sum_{Sink} J_m $$

With $m_i$ representing the mass of the isotope (i) in a given medium, and $t$ representing the time, $\dot{m}_{out} - \dot{m}_{in}$ the convection term and $J_m$ representing the material flux between 2 media (except for nuclear reactions that represent a mass transfer inside the same medium). The different transfer mechanisms taken into account in the PACTOLE V3.2 mass balance equations are indicated in Figure 3.

<figure><img src="image_4.png" /><figcaption>Figure 3: Mass transfer mechanisms integrated into the PACTOLE V3.2 code</figcaption></figure>

Corrosion of the base metal leads to the formation of an inner oxide layer on the one hand, and to the supply of ions in the primary coolant (net releases) and the formation of the outer oxide layer on the other hand. The

formation of this outer layer is partially due to the ions coming from the base metal that precipitate after mixing with the ions in the boundary layer. Therefore, radioactive ions in the coolant can be incorporated into the outer oxide layer even in a saturated medium. This modelling is based on a pragmatic approach [8] for which the parameters – especially corrosion and release rates – are determined on the basis of test loop experiments.

The dissolution of the deposit or the outer oxide layer can occur when the concentration of ions in the coolant is below the equilibrium concentration. The dissolution phenomenon is controlled by the transport kinetics of ions on the wall into the bulk of the coolant (mass transfer coefficient), by the oxide transformation kinetics, and by the concentration gradient.

Ions precipitate when their concentration in the coolant is greater than the equilibrium concentration. This occurs near a precipitation site, i.e.: near already-existing oxide, and thus near particles but especially near the wall. Ions therefore need to be transported near the wall or the particles before precipitating.

The equilibrium concentration of each chemical element is calculated with the PHREEQC code [9], in combination with a thermodynamic database developed by the CEA and coupled with PACTOLE.

This chemical module determines the composition of the ideal solid solution to model 1) a mixed oxide, 2) any pure solid phases possibly in excess (mainly metal nickel), and 3) the molar quantities of each element in equilibrium, for each control volume and in relation to the chemical conditions, the coolant temperature, and the masses of the metallic elements.

The particle deposit term is a deposit rate that takes into account:

- Turbulent diffusion and the effects of inertia [10],

- Sedimentation for horizontal ducts [11],

- Thermophoresis for temperature gradients between the coolant and the wall [12].

The erosion term involves the erosion of particles composing the deposit resulting from the coolant friction forces. The PACTOLE code model assumes that the deposit can be eroded above the laminar sub-layer.

The diffusion mechanism occurs between the inner oxide layer and the outer oxide layer, as well as between the outer oxide layer and the deposit. This mechanism follows Fick's law.

The activation or decay term describes the production or disappearance of an isotope through nuclear reactions or by radioactive cooling. Nuclear reactions occur in areas of the primary system subjected to neutron flux and are a function of the microscopic rates of these reactions.

The convection term describes the transport of particles or ions in the primary coolant through the different primary system control volumes. More specifically, this term is a function of the coolant flow rate.

Last of all, the purification term takes into account the trapping of particles and ions during their passage
## 3.2. Chemical elements and isotopes

The chemical elements taken into account in the PACTOLE V3.2 code are those composing the main alloys found in any primary system that generate the main activated corrosion products: Fe, Ni, Cr and Co.

Certain isotopes of these chemical elements become radioactive when subjected to a neutron flux. For example, the main activated corrosion products that contribute to the dose rates around the primary systems in EDF PWRs are ⁵⁸Co and ⁶⁰Co with half-lives of 70.78 days and 5.271 years respectively [7]. They are formed by the following reactions:

* ⁵⁸Ni (n,p) ⁵⁸Co

* ⁵⁹Co (n, γ) ⁶⁰Co

The mass balances are performed for each isotope of these chemical elements.

## 3.3 Mass balance and transfer mechanisms

The behaviour of the corrosion products in each control volume is modelled by the mass balance equation for each isotope in each medium [4]:

$$ \frac{\partial m_i}{\partial t} + (\dot{m}_{out} - \dot{m}_{in}) = \sum_{Source} J_m - \sum_{Sink} J_m $$

With $m_i$ representing the mass of the isotope (i) in a given medium, and $t$ representing the time, $\dot{m}_{out} - \dot{m}_{in}$ the convection term and $J_m$ representing the material flux between 2 media (except for nuclear reactions that represent a mass transfer inside the same medium). The different transfer mechanisms taken into account in the PACTOLE V3.2 mass balance equations are indicated in Figure 3.

<figure><img src="image_4.png" /><figcaption>Figure 3: Mass transfer mechanisms integrated into the PACTOLE V3.2 code</figcaption></figure>

Corrosion of the base metal leads to the formation of an inner oxide layer on the one hand, and to the supply of ions in the primary coolant (net releases) and the formation of the outer oxide layer on the other hand. The

formation of this outer layer is partially due to the ions coming from the base metal that precipitate after mixing with the ions in the boundary layer. Therefore, radioactive ions in the coolant can be incorporated into the outer oxide layer even in a saturated medium. This modelling is based on a pragmatic approach [8] for which the parameters – especially corrosion and release rates – are determined on the basis of test loop experiments.

The dissolution of the deposit or the outer oxide layer can occur when the concentration of ions in the coolant is below the equilibrium concentration. The dissolution phenomenon is controlled by the transport kinetics of ions on the wall into the bulk of the coolant (mass transfer coefficient), by the oxide transformation kinetics, and by the concentration gradient.

Ions precipitate when their concentration in the coolant is greater than the equilibrium concentration. This occurs near a precipitation site, i.e.: near already-existing oxide, and thus near particles but especially near the wall. Ions therefore need to be transported near the wall or the particles before precipitating.

The equilibrium concentration of each chemical element is calculated with the PHREEQC code [9], in combination with a thermodynamic database developed by the CEA and coupled with PACTOLE.

This chemical module determines the composition of the ideal solid solution to model 1) a mixed oxide, 2) any pure solid phases possibly in excess (mainly metal nickel), and 3) the molar quantities of each element in equilibrium, for each control volume and in relation to the chemical conditions, the coolant temperature, and the masses of the metallic elements.

The particle deposit term is a deposit rate that takes into account:

- Turbulent diffusion and the effects of inertia [10],

- Sedimentation for horizontal ducts [11],

- Thermophoresis for temperature gradients between the coolant and the wall [12].

The erosion term involves the erosion of particles composing the deposit resulting from the coolant friction forces. The PACTOLE code model assumes that the deposit can be eroded above the laminar sub-layer.

The diffusion mechanism occurs between the inner oxide layer and the outer oxide layer, as well as between the outer oxide layer and the deposit. This mechanism follows Fick's law.

The activation or decay term describes the production or disappearance of an isotope through nuclear reactions or by radioactive cooling. Nuclear reactions occur in areas of the primary system subjected to neutron flux and are a function of the microscopic rates of these reactions.

The convection term describes the transport of particles or ions in the primary coolant through the different primary system control volumes. More specifically, this term is a function of the coolant flow rate.

Last of all, the purification term takes into account the trapping of particles and ions during their passage
through the filters and demineralizers of the CVCS. This term is a function of the flow rate and the filtration efficiency.

# 4. SIMULATIONS USING PACTOLE V3.2

## 4.1. Main input data

To run the PACTOLE V3.2 code, it is necessary to provide a set of input data containing the following main data:

- The geometric parameters of a control volume are the hydraulic diameter and its wet surface.

- The thermohydraulic data such as velocity, flow and temperature of the coolant must be indicated, as well as the wall temperature.

- The materials are defined by their density, their chemical composition and their roughness. Any possible surface treatment is taken into account via the experimental corrosion and release rates.

- As the activation of the corrosion products is directly proportional to the neutron flux, the power fraction is specified in each part of the core.

- In terms of the operating data, a cycle is defined by its duration, its power history and its concentration levels in boron, lithium, hydrogen and oxygen.

## 4.2. Simulation of CRUAS-1

The 900 MWe PWR, CRUAS-1 (CPY standardized plant series), is equipped with three steam generators whose tube bundle is made of the 600 TT alloy.

Simulation of CRUAS-1 includes 196 regions, 168 of which are for the core. The core is divided into 12 sections to take into account 1/3 reloading, and then 1/4 reloading from the 4th cycle (see Figure 1).

From the 6th cycle, the fuel assembly grids made of the 718 alloy were replaced by Zircaloy grids.

The simulated primary coolant chemistry is that of the first seven cycles of CRUAS-1. CRUAS-1 operated according to a coordinated boron-lithium chemistry with a pH₃₀₀°C of 7.0 up to the third cycle, then according to a coordinated boron-lithium chemistry with a pH₃₀₀°C of 7.2. The operating parameters are given in Figure 4.

Figure 5, Figure 6 and Figure 7 compare the deposited activities for the two main radioelements (⁵⁸Co et ⁶⁰Co) calculated using PACTOLE V.3.2 with the deposited activities of the EMECC measurements [13] for the hot legs (HL), the crossover legs (CL), and the steam generator tubes (SG) (cold side).

<figure><img src="image_5.png" /><figcaption>Figure 4: CRUAS-1 – Operating parameters</figcaption></figure>

<figure><img src="image_6.png" /><figcaption>Figure 5: CRUAS-1 – Activity deposited in the hot legs</figcaption></figure>

<figure><img src="image_7.png" /><figcaption>Figure 6: CRUAS-1 – Activity deposited in the crossover legs</figcaption></figure>
through the filters and demineralizers of the CVCS. This term is a function of the flow rate and the filtration efficiency.

# 4. SIMULATIONS USING PACTOLE V3.2

## 4.1. Main input data

To run the PACTOLE V3.2 code, it is necessary to provide a set of input data containing the following main data:

- The geometric parameters of a control volume are the hydraulic diameter and its wet surface.

- The thermohydraulic data such as velocity, flow and temperature of the coolant must be indicated, as well as the wall temperature.

- The materials are defined by their density, their chemical composition and their roughness. Any possible surface treatment is taken into account via the experimental corrosion and release rates.

- As the activation of the corrosion products is directly proportional to the neutron flux, the power fraction is specified in each part of the core.

- In terms of the operating data, a cycle is defined by its duration, its power history and its concentration levels in boron, lithium, hydrogen and oxygen.

## 4.2. Simulation of CRUAS-1

The 900 MWe PWR, CRUAS-1 (CPY standardized plant series), is equipped with three steam generators whose tube bundle is made of the 600 TT alloy.

Simulation of CRUAS-1 includes 196 regions, 168 of which are for the core. The core is divided into 12 sections to take into account 1/3 reloading, and then 1/4 reloading from the 4th cycle (see Figure 1).

From the 6th cycle, the fuel assembly grids made of the 718 alloy were replaced by Zircaloy grids.

The simulated primary coolant chemistry is that of the first seven cycles of CRUAS-1. CRUAS-1 operated according to a coordinated boron-lithium chemistry with a pH₃₀₀°C of 7.0 up to the third cycle, then according to a coordinated boron-lithium chemistry with a pH₃₀₀°C of 7.2. The operating parameters are given in Figure 4.

Figure 5, Figure 6 and Figure 7 compare the deposited activities for the two main radioelements (⁵⁸Co et ⁶⁰Co) calculated using PACTOLE V.3.2 with the deposited activities of the EMECC measurements [13] for the hot legs (HL), the crossover legs (CL), and the steam generator tubes (SG) (cold side).

<figure><img src="image_5.png" /><figcaption>Figure 4: CRUAS-1 – Operating parameters</figcaption></figure>

<figure><img src="image_6.png" /><figcaption>Figure 5: CRUAS-1 – Activity deposited in the hot legs</figcaption></figure>

<figure><img src="image_7.png" /><figcaption>Figure 6: CRUAS-1 – Activity deposited in the crossover legs</figcaption></figure>
<figure><img src="image_8.png" /><figcaption>Figure 7: CRUAS-1 – Activity deposited in the steam generator tubes (cold side)</figcaption></figure>

In relation to the <sup>58</sup>Co activity measurements, the CRUAS-1 simulation overestimates the contamination in the hot legs by a factor of 2 to 3, in the crossover legs by a factor of 1.5 to 5, and in the steam generators by a factor below 1.7.

The CRUAS-1 simulation underestimates the <sup>60</sup>Co activities deposited in the hot legs by a factor of 2, in the steam generators by a factor below 1.7, and obtains an equivalent estimate in the crossover legs.

Variations in the activity levels calculated with PACTOLE V3.2 are similar to the variations in the measured activity levels:

- The deposited <sup>58</sup>Co activity tends to decrease in the legs with each cycle, and remains relatively constant in the steam generators.

- The deposited <sup>60</sup>Co activity tends to increase for several cycles.

Figure 8 compares the volume activities of <sup>58</sup>Co and <sup>60</sup>Co calculated with the PACTOLE V3.2 code with the volume activities measured by EDF.

The calculated volume activities are about several MBq/t for <sup>58</sup>Co and several tenths of MBq/t for <sup>60</sup>Co. The measured volume activities are about a dozen MBq/t for <sup>58</sup>Co and several MBq/t for <sup>60</sup>Co. The calculations therefore underestimate the volume activity of <sup>58</sup>Co and <sup>60</sup>Co by about a factor of 10 or less.

<figure><img src="image_9.png" /><figcaption>Figure 8: CRUAS-1 – Volume activity levels</figcaption></figure>

The difference between the calculated figures and the experimental measurements can therefore be significant but is greatly reduced in relation to previous versions of the code, and new improvements are expected for the future versions.

In order to validate the code and improve the modelling of activated corrosion products behaviour, comparisons between experimental results obtained with the CIRENE loop and simulations with the PACTOLE V3.2 code are also carried out [14].

Thus, the code has the advantage of a broadened validation base. One of its main advantages lies in the fact that various impact studies on one or several design and operating parameters can be carried out.

### 4.3. Influence of an operating parameter

The PACTOLE code can study the effect of design and operating parameters upon the contamination of PWR systems, e.g.: the effect of the chemistry, the composition of a material or an unscheduled shutdown during a cycle.

A shutdown with oxygenation in about the first six months of power operation can reduce the <sup>58</sup>Co concentration peak during the shutdown for refueling, and this even more true when the unscheduled shutdown with oxygenation is close to the end of the cycle [15].

To assess the impact of an unscheduled shutdown with oxygenation using the PACTOLE code, we compared the results of a simulation without an unscheduled shutdown (benchmark calculation) with simulations including an intermediary shutdown lasting 15 days and occurring 30, 60, 180 or 250 days before the end of cycle 3. The 5 simulated cycles were identical and last 300 days.

Figure 9 shows the mass of nickel deposited on the fuel that is removed by the PACTOLE code during cold shutdowns with oxygenation.
<figure><img src="image_8.png" /><figcaption>Figure 7: CRUAS-1 – Activity deposited in the steam generator tubes (cold side)</figcaption></figure>

In relation to the <sup>58</sup>Co activity measurements, the CRUAS-1 simulation overestimates the contamination in the hot legs by a factor of 2 to 3, in the crossover legs by a factor of 1.5 to 5, and in the steam generators by a factor below 1.7.

The CRUAS-1 simulation underestimates the <sup>60</sup>Co activities deposited in the hot legs by a factor of 2, in the steam generators by a factor below 1.7, and obtains an equivalent estimate in the crossover legs.

Variations in the activity levels calculated with PACTOLE V3.2 are similar to the variations in the measured activity levels:

- The deposited <sup>58</sup>Co activity tends to decrease in the legs with each cycle, and remains relatively constant in the steam generators.

- The deposited <sup>60</sup>Co activity tends to increase for several cycles.

Figure 8 compares the volume activities of <sup>58</sup>Co and <sup>60</sup>Co calculated with the PACTOLE V3.2 code with the volume activities measured by EDF.

The calculated volume activities are about several MBq/t for <sup>58</sup>Co and several tenths of MBq/t for <sup>60</sup>Co. The measured volume activities are about a dozen MBq/t for <sup>58</sup>Co and several MBq/t for <sup>60</sup>Co. The calculations therefore underestimate the volume activity of <sup>58</sup>Co and <sup>60</sup>Co by about a factor of 10 or less.

<figure><img src="image_9.png" /><figcaption>Figure 8: CRUAS-1 – Volume activity levels</figcaption></figure>

The difference between the calculated figures and the experimental measurements can therefore be significant but is greatly reduced in relation to previous versions of the code, and new improvements are expected for the future versions.

In order to validate the code and improve the modelling of activated corrosion products behaviour, comparisons between experimental results obtained with the CIRENE loop and simulations with the PACTOLE V3.2 code are also carried out [14].

Thus, the code has the advantage of a broadened validation base. One of its main advantages lies in the fact that various impact studies on one or several design and operating parameters can be carried out.

### 4.3. Influence of an operating parameter

The PACTOLE code can study the effect of design and operating parameters upon the contamination of PWR systems, e.g.: the effect of the chemistry, the composition of a material or an unscheduled shutdown during a cycle.

A shutdown with oxygenation in about the first six months of power operation can reduce the <sup>58</sup>Co concentration peak during the shutdown for refueling, and this even more true when the unscheduled shutdown with oxygenation is close to the end of the cycle [15].

To assess the impact of an unscheduled shutdown with oxygenation using the PACTOLE code, we compared the results of a simulation without an unscheduled shutdown (benchmark calculation) with simulations including an intermediary shutdown lasting 15 days and occurring 30, 60, 180 or 250 days before the end of cycle 3. The 5 simulated cycles were identical and last 300 days.

Figure 9 shows the mass of nickel deposited on the fuel that is removed by the PACTOLE code during cold shutdowns with oxygenation.
<figure><img src="image_10.png" /><figcaption>Figure 9 : Effect of an unscheduled shutdown
Mass of nickel removed during shutdowns</figcaption></figure>

It seems that the closer the unscheduled shutdown with oxygenation is to the end of the cycle, the less nickel will be removed during the end-of-cycle shutdown, and thus the weaker $ ^{58}Co $ peak will be.

Therefore, the simulations show that an unscheduled shutdown occurring 30 days before the end of the cycle reduces the mass of nickel removed by 75%, by 60% when occurring 90 days before, by 30% when occurring 180 days before, and by 15% when occurring 250 days before. By comparing these figures with feedback data [15], it was observed that the two trend lines are parallel and very similar (see Figure 10).

<figure><img src="image_11.png" /><figcaption>Figure 10 : Effect of an unscheduled shutdown
Comparison between PACTOLE and feedback</figcaption></figure>

## 5. CONCLUSION

Modelling the behaviour of corrosion products in a PWR primary system remains essential in terms of both present and future industrial stakes: radiation protection, reactor availability, environmental impact and dismantling.

The latest version of the PACTOLE code, PACTOLE V3.2, is an important step forward towards predicting PWR contamination. The main improvement is its coupling with a chemical module associated with a thermodynamic database that can be used to describe the speciation of the aqueous medium in interaction with several solid (metals, simple oxides, and mixed oxides)

and gaseous phases in reactor primary systems. Its object-oriented language makes it possible to upgrade the modelling by making relatively easy changes to equations expressing the physical conditions in question.

Comparisons between simulations performed with the PACTOLE code and measurements taken in PWRs show that variations in the activities deposited in the primary system are generally consistent with experimental feedback. The activity levels deposited for $ ^{58}Co $ and $ ^{60}Co $ – the two main contributors to dose rates around the primary system – are about the same order of magnitude as the EMECC site measurements (ratio generally below 5). In relation to the experimental values, the volume activities calculated for $ ^{58}Co $ and $ ^{60}Co $ are underestimated by a factor of 10 or less.

Simulations of the impact of certain design or operating parameters demonstrated that the code behaves consistently with experimental observations. They therefore confirm or explain the results of studies focusing on feedback.

The PACTOLE code has been developed over the past thirty years and continues to be upgraded. The next major upgrade concerns the merging of the PACTOLE code with the PROFIP† code to create the OSCAR code (French acronym for *Simulation tool of the nuclear reactor contamination*).

## 6. REFERENCES

[1] P. BESLU, G. FREJAVILLE and A. LALET, “A computer code PACTOLE to predict activation and transport of corrosion product in PWR”, Proceedings of an international conference organized by the British Nuclear Energy Society, Bournemouth, 24-27 October 1977, BNES, London, 195-201.

[2] P. BESLU and C. LEUTHROT, “PACTOLE-PROFIP : two codes allowing prediction of the contamination of PWR primary circuits”, Revue Générale Nucléaire, November-December 1990, p552-554.

[3] D. TARABELLI, S. ANTHONI, D. FÉRON, Ph. RIDOUX, L. GUINARD, C. BRUN and A. LONG, “Status and future plans of the PACTOLE code predicting the activation and transport of corrosion products in PWRs”, JAIF, International conference on water chemistry in nuclear power plants, 13-16 October 1998, Kashiwazaki (Japon), p301-305.

[4] C. MARCHETTO, D. TARABELLI, D. YOU, C. ANDRIEU, A. LONG and D. ZEITOUN, “PACTOLE V3 : a new code version to predict corrosion product contamination”, Proceedings of the International conference on water chemistry of nuclear reactor systems 8, BNES, 22-26 October 2000, Bournemouth (GB), Volume 1, p224-229.

[5] F. NGUYEN, D. TARABELLI, H. MARTEAU, F. DACQUAIT, N. PEROT, G. RANCHOUX, L. GUINARD and A. LONG, “First numerical simulations of contamination of the PWR primary circuit by

<footnote>† PROFIP : simulation code of fission product and actinide transfer in the PWR primary system [2].</footnote>
<figure><img src="image_10.png" /><figcaption>Figure 9 : Effect of an unscheduled shutdown
Mass of nickel removed during shutdowns</figcaption></figure>

It seems that the closer the unscheduled shutdown with oxygenation is to the end of the cycle, the less nickel will be removed during the end-of-cycle shutdown, and thus the weaker $ ^{58}Co $ peak will be.

Therefore, the simulations show that an unscheduled shutdown occurring 30 days before the end of the cycle reduces the mass of nickel removed by 75%, by 60% when occurring 90 days before, by 30% when occurring 180 days before, and by 15% when occurring 250 days before. By comparing these figures with feedback data [15], it was observed that the two trend lines are parallel and very similar (see Figure 10).

<figure><img src="image_11.png" /><figcaption>Figure 10 : Effect of an unscheduled shutdown
Comparison between PACTOLE and feedback</figcaption></figure>

## 5. CONCLUSION

Modelling the behaviour of corrosion products in a PWR primary system remains essential in terms of both present and future industrial stakes: radiation protection, reactor availability, environmental impact and dismantling.

The latest version of the PACTOLE code, PACTOLE V3.2, is an important step forward towards predicting PWR contamination. The main improvement is its coupling with a chemical module associated with a thermodynamic database that can be used to describe the speciation of the aqueous medium in interaction with several solid (metals, simple oxides, and mixed oxides)

and gaseous phases in reactor primary systems. Its object-oriented language makes it possible to upgrade the modelling by making relatively easy changes to equations expressing the physical conditions in question.

Comparisons between simulations performed with the PACTOLE code and measurements taken in PWRs show that variations in the activities deposited in the primary system are generally consistent with experimental feedback. The activity levels deposited for $ ^{58}Co $ and $ ^{60}Co $ – the two main contributors to dose rates around the primary system – are about the same order of magnitude as the EMECC site measurements (ratio generally below 5). In relation to the experimental values, the volume activities calculated for $ ^{58}Co $ and $ ^{60}Co $ are underestimated by a factor of 10 or less.

Simulations of the impact of certain design or operating parameters demonstrated that the code behaves consistently with experimental observations. They therefore confirm or explain the results of studies focusing on feedback.

The PACTOLE code has been developed over the past thirty years and continues to be upgraded. The next major upgrade concerns the merging of the PACTOLE code with the PROFIP† code to create the OSCAR code (French acronym for *Simulation tool of the nuclear reactor contamination*).

## 6. REFERENCES

[1] P. BESLU, G. FREJAVILLE and A. LALET, “A computer code PACTOLE to predict activation and transport of corrosion product in PWR”, Proceedings of an international conference organized by the British Nuclear Energy Society, Bournemouth, 24-27 October 1977, BNES, London, 195-201.

[2] P. BESLU and C. LEUTHROT, “PACTOLE-PROFIP : two codes allowing prediction of the contamination of PWR primary circuits”, Revue Générale Nucléaire, November-December 1990, p552-554.

[3] D. TARABELLI, S. ANTHONI, D. FÉRON, Ph. RIDOUX, L. GUINARD, C. BRUN and A. LONG, “Status and future plans of the PACTOLE code predicting the activation and transport of corrosion products in PWRs”, JAIF, International conference on water chemistry in nuclear power plants, 13-16 October 1998, Kashiwazaki (Japon), p301-305.

[4] C. MARCHETTO, D. TARABELLI, D. YOU, C. ANDRIEU, A. LONG and D. ZEITOUN, “PACTOLE V3 : a new code version to predict corrosion product contamination”, Proceedings of the International conference on water chemistry of nuclear reactor systems 8, BNES, 22-26 October 2000, Bournemouth (GB), Volume 1, p224-229.

[5] F. NGUYEN, D. TARABELLI, H. MARTEAU, F. DACQUAIT, N. PEROT, G. RANCHOUX, L. GUINARD and A. LONG, “First numerical simulations of contamination of the PWR primary circuit by

<footnote>† PROFIP : simulation code of fission product and actinide transfer in the PWR primary system [2].</footnote>
activated corrosion products with the PACTOLE V3.0 code", Proceedings of the International conference on Water chemistry in nuclear reactors systems, 11-14 October 2004, San Francisco (US), Poster ½.

[6] D.H. LISTER and R.D. DAVIDSON, "Corrosion-Product Release in Light Water Reactors", NP-6512 Research Project 2008-1, Final report, September 1989.

[7] F. DACQUAIT, C. ANDRIEU, L. GUINARD, J.L. BRETELLE, F. BARDET, A. ROCHER and C. BRUN, "Status of primary system contamination in French PWRs", Proceedings of the International conference on Water chemistry in nuclear reactors systems, 11-14/10/2004, San Francisco (USA), Session 2 Paper 4.

[8] F. COULET, «Etude et modélisation de l'influence des matériaux sur la contamination du circuit primaire des réacteurs à eau pressurisée par les produits de corrosion », Thèse de doctorat de l'INPG, 1996, Grenoble (France).

[9] D.L. PARKHURST and C.A.J. APPELO, "User's guide to PHREEQC (version 2) – A computer program for speciation, batch-reactions, one-dimensional transport, and inverse geochemical calculations", Water-Ressource Investigations Report 99-4259, 1999.

[10] S.K BEAL, "Deposition of particles in turbulent flow on channel or pipe walls", Nuclear science and engineering, 1970, vol. 40, p.1-11.

[11] S.J. MICHELL, "Fluid and particle mechanics", Pergamon press Ltd, 1970.

[12] A.C. PONTING and R.S. RODLIFE, "Intrinsic filtration and retarded deposition for the control of colloidal corrosion product deposition on PWR fuel", Water chemistry of nuclear reactor system, Proceedings of the BNES, 1983, p. 43-51, Bournemouth UK.

[13] R. EIMECKE, S. ANTHONI, «Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC)», 7th International Conference on Radiation Shielding, 12-15 September 1988, Bournemouth (England).

[14] M. GIRARD, F. DACQUAIT, H. MARTEAU, B. LARAT, F. NGUYEN, L. GUINARD and C. VIALA, "The CIRENE loop: a tool to study ACP deposits and to validate the PACTOLE code", this conference.

[15] F. DACQUAIT, C. ANDRIEU, M. BERGER, J.L. BRETELLE and A. ROCHER, "Corrosion product transfer in French PWRs during shutdown", Proceedings of the International Conference on Water Chemistry in Nuclear Reactor systems, 22-26 April 2002, Avignon (France), papier 113.
activated corrosion products with the PACTOLE V3.0 code", Proceedings of the International conference on Water chemistry in nuclear reactors systems, 11-14 October 2004, San Francisco (US), Poster ½.

[6] D.H. LISTER and R.D. DAVIDSON, "Corrosion-Product Release in Light Water Reactors", NP-6512 Research Project 2008-1, Final report, September 1989.

[7] F. DACQUAIT, C. ANDRIEU, L. GUINARD, J.L. BRETELLE, F. BARDET, A. ROCHER and C. BRUN, "Status of primary system contamination in French PWRs", Proceedings of the International conference on Water chemistry in nuclear reactors systems, 11-14/10/2004, San Francisco (USA), Session 2 Paper 4.

[8] F. COULET, «Etude et modélisation de l'influence des matériaux sur la contamination du circuit primaire des réacteurs à eau pressurisée par les produits de corrosion », Thèse de doctorat de l'INPG, 1996, Grenoble (France).

[9] D.L. PARKHURST and C.A.J. APPELO, "User's guide to PHREEQC (version 2) – A computer program for speciation, batch-reactions, one-dimensional transport, and inverse geochemical calculations", Water-Ressource Investigations Report 99-4259, 1999.

[10] S.K BEAL, "Deposition of particles in turbulent flow on channel or pipe walls", Nuclear science and engineering, 1970, vol. 40, p.1-11.

[11] S.J. MICHELL, "Fluid and particle mechanics", Pergamon press Ltd, 1970.

[12] A.C. PONTING and R.S. RODLIFE, "Intrinsic filtration and retarded deposition for the control of colloidal corrosion product deposition on PWR fuel", Water chemistry of nuclear reactor system, Proceedings of the BNES, 1983, p. 43-51, Bournemouth UK.

[13] R. EIMECKE, S. ANTHONI, «Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC)», 7th International Conference on Radiation Shielding, 12-15 September 1988, Bournemouth (England).

[14] M. GIRARD, F. DACQUAIT, H. MARTEAU, B. LARAT, F. NGUYEN, L. GUINARD and C. VIALA, "The CIRENE loop: a tool to study ACP deposits and to validate the PACTOLE code", this conference.

[15] F. DACQUAIT, C. ANDRIEU, M. BERGER, J.L. BRETELLE and A. ROCHER, "Corrosion product transfer in French PWRs during shutdown", Proceedings of the International Conference on Water Chemistry in Nuclear Reactor systems, 22-26 April 2002, Avignon (France), papier 113.