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

Thomas Jobert, Sophie Raphaël, Denis Pombet, Frédéric Dacquait, Lucie d'Ascenzo, Ludovic Vaillant, Inès Dignito, Thomas Chirent

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Thomas Jobert, Sophie Raphaël, Denis Pombet, Frédéric Dacquait, Lucie d'Ascenzo, et al.. New status of the primary system contamination of the French PWRs - An updated analysis of the EMECC data. NPC 2023 - International Conference on Nuclear Plant Chemistry - International Conference on Water Chemistry in Nuclear Reactor Systems, SFEN, Sep 2023, Juan-les-Pins, France. cea-04475843

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# N°000140 New Status of the Primary System Contamination of the French PWRs – An updated analysis of the EMECC data

Thomas Jobert<fnref n="1" />*, Sophie Raphaël<fnref n="1" />, Denis Pombet<fnref n="2" />, Frédéric Dacquait<fnref n="3" />, Lucie D'Ascenzo<fnref n="4" />, Ludovic Vaillant<fnref n="4" />, Inès Dignito<fnref n="5" />, Thomas Chirent<fnref n="1" />

## Abstract

This paper presents the “new status of the reactor cooling system (RCS) contamination of the French fleet”, considering the updated EMECC measurement database. The analysis focuses on radiocobalts Co-58 and Co-60, the surface activities of which are responsible of around 85% of the dose received by workers during shutdown and maintenance operations in PWRs. The main objectives of the study are to identify whether the contamination trends described 15 years ago are still relevant or not, and to assess the impact of the following operating practices or design options on deposited activities: steam generator replacement (SGR), zinc injection, cold shutdown procedure, steam generator (SG) tube manufacturing processes. The main lessons learned are that after SGR the Co-58 contamination in the RCS reaches very low levels on the 900 MWe series whereas, in the long-term, Co-60 contamination seems to be increasing again after the decrease trend described in 2005. EMECC measurements tend to confirm that zinc injection is likely to reduce Co-58 surface contamination, providing injection has started before the formation of the passive oxide layer. Lastly, EMECC measurements also suggest a positive effect of the recent evolutions introduced in the SG tube manufacturing process on Co-58 surface contamination.

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

## Introduction

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

### 1. EMECC measurement campaigns

The EMECC campaigns consist in measuring the activities deposited inside some of the components (pipes, steam generator tubing, heat exchangers...) of the primary and auxiliary systems. The surface activities are measured by gamma spectrometry measurements using the so-called EMECC device designed by the CEA ([1]) and equipped with a High Purity Germanium detector. The background noise and the volume activities are taken into account for each EMECC measurement in order to determine the surface activities only ([2]).

Surface activities inside primary circuits of the French fleet are thus collected since 1971.

As explained in [3], complementarily to dose rate measurements performed at each outage (around primary and auxiliary circuits), EDF selects units, in the frame of a plurennial program, on which EMECC measurement campaigns are conducted. Currently, about 8 EMECC measurement campaigns are carried out every year amongst the 3 reactor series of the French fleet.

Through EMECC measurements, EDF is able to gather accurate data on the surface contamination of each reactor of the fleet and can improve its understanding of contamination mechanisms. The surface activities spectrum obtained with the EMECC device is typically composed of the following radionuclides: Co-58, Co-60, Cr-51, Ag-110m, Sb-124, Fe-59, Mn-54, Zr-95, Nb-95, Zn-65. In addition

<footnote n="1">¹ EDF/DIPNN/DT, F-69363 Lyon, France.</footnote>

<footnote n="2">² EDF/DIPNN/DI, Paris, France.</footnote>

<footnote n="3">³ CEA, DES, IRESNE, DTN, Cadarache, F-13108 Saint-Paul-Lez-Durance, France.</footnote>

<footnote n="4">⁴ CEPN, France.</footnote>

<footnote n="5">⁵ EDVANCE, Lyon, France.</footnote>

<footnote>thomas.jobert@edf.fr</footnote>
# N°000140 New Status of the Primary System Contamination of the French PWRs – An updated analysis of the EMECC data

Thomas Jobert¹*, Sophie Raphaël¹, Denis Pombet², Frédéric Dacquait³, Lucie D'Ascenzo⁴, Ludovic Vaillant⁴, Inès Dignito⁵, Thomas Chirent¹

¹ EDF/DIPNN/DT, F-69363 Lyon, France.

² EDF/DIPNN/DI, Paris, France.

³ CEA, DES, IRESNE, DTN, Cadarache, F-13108 Saint-Paul-Lez-Durance, France.

⁴ CEPN, France.

⁵ EDVANCE, Lyon, France.

thomas.jobert@edf.fr

## Abstract

This paper presents the “new status of the reactor cooling system (RCS) contamination of the French fleet”, considering the updated EMECC measurement database. The analysis focuses on radiocobalts Co-58 and Co-60, the surface activities of which are responsible of around 85% of the dose received by workers during shutdown and maintenance operations in PWRs. The main objectives of the study are to identify whether the contamination trends described 15 years ago are still relevant or not, and to assess the impact of the following operating practices or design options on deposited activities: steam generator replacement (SGR), zinc injection, cold shutdown procedure, steam generator (SG) tube manufacturing processes. The main lessons learned are that after SGR the Co-58 contamination in the RCS reaches very low levels on the 900 MWe series whereas, in the long-term, Co-60 contamination seems to be increasing again after the decrease trend described in 2005. EMECC measurements tend to confirm that zinc injection is likely to reduce Co-58 surface contamination, providing injection has started before the formation of the passive oxide layer. Lastly, EMECC measurements also suggest a positive effect of the recent evolutions introduced in the SG tube manufacturing process on Co-58 surface contamination.

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

## Introduction

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

### 1. EMECC measurement campaigns

The EMECC campaigns consist in measuring the activities deposited inside some of the components (pipes, steam generator tubing, heat exchangers...) of the primary and auxiliary systems. The surface activities are measured by gamma spectrometry measurements using the so-called EMECC device designed by the CEA ([1]) and equipped with a High Purity Germanium detector. The background noise and the volume activities are taken into account for each EMECC measurement in order to determine the surface activities only ([2]).

Surface activities inside primary circuits of the French fleet are thus collected since 1971.

As explained in [3], complementarily to dose rate measurements performed at each outage (around primary and auxiliary circuits), EDF selects units, in the frame of a plurennial program, on which EMECC measurement campaigns are conducted. Currently, about 8 EMECC measurement campaigns are carried out every year amongst the 3 reactor series of the French fleet.

Through EMECC measurements, EDF is able to gather accurate data on the surface contamination of each reactor of the fleet and can improve its understanding of contamination mechanisms. The surface activities spectrum obtained with the EMECC device is typically composed of the following radionuclides: Co-58, Co-60, Cr-51, Ag-110m, Sb-124, Fe-59, Mn-54, Zr-95, Nb-95, Zn-65. In addition
to these activated corrosion products, fission products, such as Cs-134, Cs-137, I-131, Te-132, are sometimes detected.

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

- On the primary circuit: primary coolant pipes (hot, crossover and cold legs) and Steam Generator (SG) tube bundle of the primary system loops,

- On the auxiliary circuits: heat exchangers and pipes of the Chemical and Volume Control System (CVCS), the Residual Heat Removal System (RHRS), the Nuclear Sampling System (NSS)...

The main objectives of the EMECC measurements are:

- To assess the impact of operating practices or parameters on the surface contamination, and indirectly on the radiation field (impact of dissolved H₂ content, zinc injection, B/Li coordination...).

- To understand contamination mechanisms and elaborate strategies to cope with said mechanisms, either by preventive or curative means. This includes shutdown strategies, as EMECC measurements are performed before and after oxygenation,

- To identify the origins of pollution and develop means (material substitutions, design changes, ...) to limit the productions of detrimental radionuclides,

- To develop, calibrate and validate calculation tools to model and predict the contamination phenomena: the CEA has developed, in the frame of a tripartite Institute (CEA-EDF-Framatome), the OSCAR code, which latest version is the V1.4a ([4]).

- To gather a sufficient amount of data in order to launch data science-based analyses in the near future.

## 2. French PWR contamination

In this paper, we will focus on radiocobalts Co-58 and Co-60, which are the main contributors to the dose rates around the PWR primary system and the dosimetry on the French fleet. As explained in the introduction, one of the objectives of the work presented here is to compare the current trend of the RCS contamination to the previous results displayed in [3], which also focused on these 2 radionuclides.

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

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

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

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

- Before steam generator replacements (SGR),

- After SGR.
to these activated corrosion products, fission products, such as Cs-134, Cs-137, I-131, Te-132, are sometimes detected.

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

- On the primary circuit: primary coolant pipes (hot, crossover and cold legs) and Steam Generator (SG) tube bundle of the primary system loops,

- On the auxiliary circuits: heat exchangers and pipes of the Chemical and Volume Control System (CVCS), the Residual Heat Removal System (RHRS), the Nuclear Sampling System (NSS)...

The main objectives of the EMECC measurements are:

- To assess the impact of operating practices or parameters on the surface contamination, and indirectly on the radiation field (impact of dissolved H₂ content, zinc injection, B/Li coordination...).

- To understand contamination mechanisms and elaborate strategies to cope with said mechanisms, either by preventive or curative means. This includes shutdown strategies, as EMECC measurements are performed before and after oxygenation,

- To identify the origins of pollution and develop means (material substitutions, design changes, ...) to limit the productions of detrimental radionuclides,

- To develop, calibrate and validate calculation tools to model and predict the contamination phenomena: the CEA has developed, in the frame of a tripartite Institute (CEA-EDF-Framatome), the OSCAR code, which latest version is the V1.4a ([4]).

- To gather a sufficient amount of data in order to launch data science-based analyses in the near future.

## 2. French PWR contamination

In this paper, we will focus on radiocobalts Co-58 and Co-60, which are the main contributors to the dose rates around the PWR primary system and the dosimetry on the French fleet. As explained in the introduction, one of the objectives of the work presented here is to compare the current trend of the RCS contamination to the previous results displayed in [3], which also focused on these 2 radionuclides.

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

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

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

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

- Before steam generator replacements (SGR),

- After SGR.
## 2.1. Co-58 Contamination

### 2.1.1. Before SGR

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

**Primary pipes:**

<figure><img src="image_1.png" /><figcaption>Figure 1: Before SGR - Evolution of Co-58 surface activities in the HL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_2.png" /><figcaption>Figure 2: Before SGR - Evolution of Co-58 surface activities in the CL and COL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>
## 2.1. Co-58 Contamination

### 2.1.1. Before SGR

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

**Primary pipes:**

<figure><img src="image_2.png" /><figcaption>Figure 1: Before SGR - Evolution of Co-58 surface activities in the HL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_3.png" /><figcaption>Figure 2: Before SGR - Evolution of Co-58 surface activities in the CL and COL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>
Steam generators:

<figure><img src="image_3.png" /><figcaption>Figure 3: Before SGR – Evolution of Co-58 surface activities in the SG tubing (cold part) of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_4.png" /></figure>

The update of the data confirms that the Co-58 deposited activity is all the lower as the series is newer all along the period considered, 1450 MWe being the most recent series and 900 MWe the older one (except for the hot legs, where the gap seems to be reduced with time). Similarly, the contamination range narrowing with reactor types is confirmed when a longer period of measurements is available. Several arguments have been proposed to explain this discrepancy among the 3 series:

- pH₃₀₀°C: the first reactors operated with a low pH, before implementing a boron/lithium coordination with a target pH₃₀₀°C fixed at 7.2. But as mentioned in [5][6] and reminded in [3], the impact of the pH on surface contamination is difficult to estimate from OPEX data analyses only. In paper [7], calculations have been performed with the OSCAR code to assess the effect of pH only on the surface contamination: the results showed a moderated impact of the pH, leading, for example, to a decrease of around 5 % of the Co-58 deposited activities in the SG when moving from a lithium content of 2.2 ppm to 3.5 ppm, in order to increase the pH₃₀₀°C = 7.2 duration over the cycle.

- SG tube alloys and SG tube manufacturing processes (see [8], [9]): The French PWR SG are equipped with tubing made of 600 Mill-Annealed (MA) alloy for the oldest, 600 Thermally-Treated (TT) or 690 TT for the most recent ones. The impact of the SG alloys can also be identified when considering EMECC measurements after SGR (see below: §2.1.2). In this paper, the impact of the manufacturing process of 690TT alloy SG tubes on the RCS surface contamination is assessed in §5.

For the primary pipes, whatever the series selected, the global trend of the curves is similar: after a sharp decrease of the Co-58 surface contamination within 6 to 8 cycles, followed by a smoother one, an almost stable state is reached. This evolution is consistent with the nickel release rate of SG tubes measured in loop tests. The plateau is reached sooner for the hot legs, compared to the cold legs. It can also be noticed that the dispersion of the values is globally decreasing given the cycles. This general trend is directly linked to the formation of the passive protective oxide layer on the surface of the SG tubes, decreasing the nickel release from the tubing over time. The replacement of the spacer grids, initially made in Inconel 718, by grids made in Zircaloy alloy, has also contributed to the reduction of the Co-58 surface contamination, as already explained in [3]: the former spacer grids of the assemblies being composed mainly by nickel and submitted to the neutron flux, they likely constituted a non-negligible source of Co-58, whereas Ni can be present as impurity only in Zircaloy-made materials.

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

<figure><img src="image_4.png" /><figcaption>Figure 3: Before SGR – Evolution of Co-58 surface activities in the SG tubing (cold part) of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_5.png" /></figure>

The update of the data confirms that the Co-58 deposited activity is all the lower as the series is newer all along the period considered, 1450 MWe being the most recent series and 900 MWe the older one (except for the hot legs, where the gap seems to be reduced with time). Similarly, the contamination range narrowing with reactor types is confirmed when a longer period of measurements is available. Several arguments have been proposed to explain this discrepancy among the 3 series:

- pH₃₀₀°C: the first reactors operated with a low pH, before implementing a boron/lithium coordination with a target pH₃₀₀°C fixed at 7.2. But as mentioned in [5][6] and reminded in [3], the impact of the pH on surface contamination is difficult to estimate from OPEX data analyses only. In paper [7], calculations have been performed with the OSCAR code to assess the effect of pH only on the surface contamination: the results showed a moderated impact of the pH, leading, for example, to a decrease of around 5 % of the Co-58 deposited activities in the SG when moving from a lithium content of 2.2 ppm to 3.5 ppm, in order to increase the pH₃₀₀°C = 7.2 duration over the cycle.

- SG tube alloys and SG tube manufacturing processes (see [8], [9]): The French PWR SG are equipped with tubing made of 600 Mill-Annealed (MA) alloy for the oldest, 600 Thermally-Treated (TT) or 690 TT for the most recent ones. The impact of the SG alloys can also be identified when considering EMECC measurements after SGR (see below: §2.1.2). In this paper, the impact of the manufacturing process of 690TT alloy SG tubes on the RCS surface contamination is assessed in §5.

For the primary pipes, whatever the series selected, the global trend of the curves is similar: after a sharp decrease of the Co-58 surface contamination within 6 to 8 cycles, followed by a smoother one, an almost stable state is reached. This evolution is consistent with the nickel release rate of SG tubes measured in loop tests. The plateau is reached sooner for the hot legs, compared to the cold legs. It can also be noticed that the dispersion of the values is globally decreasing given the cycles. This general trend is directly linked to the formation of the passive protective oxide layer on the surface of the SG tubes, decreasing the nickel release from the tubing over time. The replacement of the spacer grids, initially made in Inconel 718, by grids made in Zircaloy alloy, has also contributed to the reduction of the Co-58 surface contamination, as already explained in [3]: the former spacer grids of the assemblies being composed mainly by nickel and submitted to the neutron flux, they likely constituted a non-negligible source of Co-58, whereas Ni can be present as impurity only in Zircaloy-made materials.

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

SG with 600 alloy tubes of the 900 MWe series have progressively been replaced by SG equipped with 690 alloy tubing since the 1990s.

Figure 4 and Figure 5 present the mean Co-58 deposited activities after SGR, respectively in the HL, in the CL (or COL) and in the SG tubing of 900 MWe PWRs. Note that the first value displayed on each graphic corresponds to the state before SGR, in order to facilitate the comparison between the surface contamination before and after SGR. This single value before SGR represents the average of the contamination measured for the 6 cycles preceding the SGR.

**Primary pipes:**

<figure><img src="image_5.png" /><figcaption>Figure 4: After SGR – Evolution of Co-58 surface activities in HL (left) and CL&COL (right) of the 900 MWe series of the French fleet</figcaption></figure>

**SG:**

<figure><img src="image_6.png" /><figcaption>Figure 5: After SGR – Evolution of Co-58 surface activities in the SG tubing (hot and cold parts) of the 900 MWe series of the French fleet</figcaption></figure>

For primary pipes, after SGR, the global evolution of the Co-58 surface contamination is similar to that described in § 2.1.1, but the magnitude of the contamination is far much lower than observed with 600 alloy SG tubes and close to the level obtained with the most recent reactors (1450 MWe series). After 7 to 8 cycles, the Co-58 surface contamination is lower than the surface contamination state before SG. Although changes to limit the contamination have been introduced since the start-up of the 900 MWe plants, as mentioned above, the decrease of the Co-58 surface contamination is undoubtedly due to the new SG (decrease of the Ni content into the alloy and manufacturing process improvement – see [8] and [9]).

Note that after the SGR, for the first cycles, the contamination is higher in the cold legs, compared to the hot legs, as it was before SGR.

For SG tubing, after their replacement, the Co-58 average surface contamination fluctuates for 6 to 12 cycles (as observed before SGR in §2.1.1 for the 900 and 1300 MWe series), depending on the SG
### 2.1.2. After SGR

SG with 600 alloy tubes of the 900 MWe series have progressively been replaced by SG equipped with 690 alloy tubing since the 1990s.

Figure 4 and Figure 5 present the mean Co-58 deposited activities after SGR, respectively in the HL, in the CL (or COL) and in the SG tubing of 900 MWe PWRs. Note that the first value displayed on each graphic corresponds to the state before SGR, in order to facilitate the comparison between the surface contamination before and after SGR. This single value before SGR represents the average of the contamination measured for the 6 cycles preceding the SGR.

**Primary pipes:**

<figure><img src="image_6.png" /><figcaption>Figure 4: After SGR – Evolution of Co-58 surface activities in HL (left) and CL&COL (right) of the 900 MWe series of the French fleet</figcaption></figure>

**SG:**

<figure><img src="image_7.png" /><figcaption>Figure 5: After SGR – Evolution of Co-58 surface activities in the SG tubing (hot and cold parts) of the 900 MWe series of the French fleet</figcaption></figure>

For primary pipes, after SGR, the global evolution of the Co-58 surface contamination is similar to that described in § 2.1.1, but the magnitude of the contamination is far much lower than observed with 600 alloy SG tubes and close to the level obtained with the most recent reactors (1450 MWe series). After 7 to 8 cycles, the Co-58 surface contamination is lower than the surface contamination state before SG. Although changes to limit the contamination have been introduced since the start-up of the 900 MWe plants, as mentioned above, the decrease of the Co-58 surface contamination is undoubtedly due to the new SG (decrease of the Ni content into the alloy and manufacturing process improvement – see [8] and [9]).

Note that after the SGR, for the first cycles, the contamination is higher in the cold legs, compared to the hot legs, as it was before SGR.

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

## 2.2. Co-60 contamination

### 2.2.1. Before SGR

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

**Primary pipes:**

<figure><img src="image_7.png" /></figure>

<figure><img src="image_8.png" /></figure>

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

<figure><img src="image_9.png" /></figure>

<figure><img src="image_10.png" /></figure>

*Figure 7: Before SGR - Evolution of Co-60 surface activities in the CL and COL of the French fleet given the reactor series (maximum/minimum and average values)*
subregions, with activities in the same order of magnitude as just before the SGR. Then, the Co-58 surface contamination reaches low values, similar to the 1450 MWe series which operates with 690 alloy tubing since the start-up.

## 2.2. Co-60 contamination

### 2.2.1. Before SGR

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

**Primary pipes:**

<figure><img src="image_8.png" /><figcaption>Figure 6: Before SGR - Evolution of Co-60 surface activities in the HL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_9.png" /><figcaption>Figure 7: Before SGR – Evolution of Co-60 surface activities in the CL and COL of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>
SG:

<figure><img src="image_11.png" /><figcaption>Figure 8: Before SGR – Evolution of Co-60 surface activities in the SG tubing (cold part) of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_12.png" /></figure>

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

On the other hand, for cold legs, an increase of the surface contamination for 4 to 10 cycles (depending on the series) is observed followed by a slow decrease over the cycles.

Contrarily to the Co-58 situation (see §2.1.1), the discrepancy between reactor types is not as marked: the Co-60 surface contamination of the most recent reactors is not systematically lower than those of the 1300 MWe series. This observation is new and is in opposition of what could be expected from the first measurements on the 1450 MWe series (see [3]).

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

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

The SG tubings are all the more contaminated in Co-60 as the reactor series is old.

### 2.2.2. After SGR

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

<figure><img src="image_10.png" /><figcaption>Figure 8: Before SGR – Evolution of Co-60 surface activities in the SG tubing (cold part) of the French fleet given the reactor series (maximum/minimum and average values)</figcaption></figure>

<figure><img src="image_11.png" /></figure>

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

On the other hand, for cold legs, an increase of the surface contamination for 4 to 10 cycles (depending on the series) is observed followed by a slow decrease over the cycles.

Contrarily to the Co-58 situation (see §2.1.1), the discrepancy between reactor types is not as marked: the Co-60 surface contamination of the most recent reactors is not systematically lower than those of the 1300 MWe series. This observation is new and is in opposition of what could be expected from the first measurements on the 1450 MWe series (see [3]).

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

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

The SG tubings are all the more contaminated in Co-60 as the reactor series is old.

### 2.2.2. After SGR

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

<figure><img src="image_13.png" /></figure>

<figure><img src="image_14.png" /></figure>

Figure 9: After SGR – Evolution of Co-60 surface activities in HL (left) and CL&COL (right) of the 900 MWe series of the French fleet

**SG:**

<figure><img src="image_15.png" /></figure>

<figure><img src="image_16.png" /></figure>

Figure 10: After SGR – Evolution of Co-60 surface activities in the SG tubing (hot and cold parts) of the 900 MWe series of the French fleet

Immediately after SGR, the Co-60 surface contamination is lower than the reference state before SGR in all parts of the RCS. Then, a steady increase of the contamination is measured:

- Into the primary pipes, this increase is observed later in the cold legs than in the hot legs. In these two regions, the Co-60 deposited activities overpass that of before SGR: in the short-term (4 cycles) for the hot legs, in the long-term (18 cycles) for the cold legs.

- Into the SG tubing, this increase is slow and does not reach the reference level before SGR, even after 20 cycles.

Several origins of Co-60 surface contamination have already long been identified:

- Wear of stellites surfaces used in valves, bearing and stops of pumps, control rod driven mechanisms (CRDM), core internals. The corrosion of stellites surfaces also contributes to Co-60 contamination but these releases are usually low compared to other Co-59 related sources,

- Corrosion and release from alloys, especially from SG tubes (even if Co-59 is present as impurities in the alloy, the total wet surface of the tubing leads to a significant contribution to the Co-60 surface contamination of the circuits),

- Release of activated in core materials,

- Unplanned introduction of Co-59 in the primary coolant while undertaking maintenance operations during outage (for example, when grinding seat valves to avoid leakage),

- Operating events (bearing degradation, for example).

As explained in [3], the answer to the question of the main contributor to Co-60 surface contamination depends on the available source term of Co-59 (surface of stellites, content of Co-59 impurities in the alloys ...) and its release rate (normal wear, unexpected events ...), which is also linked to operating
Primary pipes:

<figure><img src="image_12.png" /></figure>

<figure><img src="image_13.png" /></figure>

Figure 9: After SGR – Evolution of Co-60 surface activities in HL (left) and CL&COL (right) of the 900 MWe series of the French fleet

SG:

<figure><img src="image_14.png" /></figure>

<figure><img src="image_15.png" /></figure>

Figure 10: After SGR – Evolution of Co-60 surface activities in the SG tubing (hot and cold parts) of the 900 MWe series of the French fleet

Immediately after SGR, the Co-60 surface contamination is lower than the reference state before SGR in all parts of the RCS. Then, a steady increase of the contamination is measured:

- Into the primary pipes, this increase is observed later in the cold legs than in the hot legs. In these two regions, the Co-60 deposited activities overpass that of before SGR: in the short-term (4 cycles) for the hot legs, in the long-term (18 cycles) for the cold legs.

- Into the SG tubing, this increase is slow and does not reach the reference level before SGR, even after 20 cycles.

Several origins of Co-60 surface contamination have already long been identified:

- Wear of stellites surfaces used in valves, bearing and stops of pumps, control rod driven mechanisms (CRDM), core internals. The corrosion of stellites surfaces also contributes to Co-60 contamination but these releases are usually low compared to other Co-59 related sources,

- Corrosion and release from alloys, especially from SG tubes (even if Co-59 is present as impurities in the alloy, the total wet surface of the tubing leads to a significant contribution to the Co-60 surface contamination of the circuits),

- Release of activated in core materials,

- Unplanned introduction of Co-59 in the primary coolant while undertaking maintenance operations during outage (for example, when grinding seat valves to avoid leakage),

- Operating events (bearing degradation, for example).

As explained in [3], the answer to the question of the main contributor to Co-60 surface contamination depends on the available source term of Co-59 (surface of stellites, content of Co-59 impurities in the alloys ...) and its release rate (normal wear, unexpected events ...), which is also linked to operating
conditions (pH, dissolved H₂ content ...) or design options (CRDM load ...). The sources being multiple, the main one depends on the level of the others.

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

- For the primary pipes: Less variations of the values from one cycle to the next occurred after SGR than before; and when we limit the comparison of the Co-60 surface activities during the 15 first cycles of each period, the activity measured after SGR remains below that of before SGR.

- For the SG tubing: 20 cycles after SGR, the Co-60 surface contamination is twice lower than the level measured just before SGR, in spite of the steady increase observed.

Regarding radiation field concerns, the SGR performed on the 900 MWe reactor series lead to a positive impact, by reducing the Co-60 surface contamination level in SG in a sustainable manner. Prior to the SGR, (one of) the main source of Co-60 surface contamination in the 900 MWe series was probably the cobalt impurities of the SG in 600 alloy.

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

### 3. Zinc injection

One of the key questions related to primary surface contamination in paper [3] was the effect of zinc injection, which was about to be tested in 2004, and for which beneficial effects were expected, especially regarding Co-60 contamination.

Since then, zinc injection has been widely implemented in reactors of the French fleet and a first analysis of the EMECC measurement campaigns conducted in France has been proposed [10].

In PWRs, zinc is injected mainly to:

- reduce stress corrosion cracking, especially of 600 MA alloy (material aspects),

- lower the radiocobalts source term, decrease the surface contamination in Co-58 and Co-60 and diminish the dose rates (source term and radiation protection aspects), and

- limit and homogenise the deposit on the core and prevent axial offset anomaly (AOA) phenomena (fuel aspects).

Nevertheless, the positive impact of zinc injection on radiation field is still on debate at the international level.

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

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

- the curative mode: zinc is injected after SGR (or after the start-up of a new reactor).

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

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

Figure 11 to Figure 14 present Co-58 and Co-60 surface activities, for different sub-regions of the RCS, for each injection mode. Data related to units without (respectively with) zinc injection are in blue (respectively orange).
conditions (pH, dissolved H₂ content ...) or design options (CRDM load ...). The sources being multiple, the main one depends on the level of the others.

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

- For the primary pipes: Less variations of the values from one cycle to the next occurred after SGR than before; and when we limit the comparison of the Co-60 surface activities during the 15 first cycles of each period, the activity measured after SGR remains below that of before SGR.

- For the SG tubing: 20 cycles after SGR, the Co-60 surface contamination is twice lower than the level measured just before SGR, in spite of the steady increase observed.

Regarding radiation field concerns, the SGR performed on the 900 MWe reactor series lead to a positive impact, by reducing the Co-60 surface contamination level in SG in a sustainable manner. Prior to the SGR, (one of) the main source of Co-60 surface contamination in the 900 MWe series was probably the cobalt impurities of the SG in 600 alloy.

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

### 3. Zinc injection

One of the key questions related to primary surface contamination in paper [3] was the effect of zinc injection, which was about to be tested in 2004, and for which beneficial effects were expected, especially regarding Co-60 contamination.

Since then, zinc injection has been widely implemented in reactors of the French fleet and a first analysis of the EMECC measurement campaigns conducted in France has been proposed [10].

In PWRs, zinc is injected mainly to:

- reduce stress corrosion cracking, especially of 600 MA alloy (material aspects),

- lower the radiocobalts source term, decrease the surface contamination in Co-58 and Co-60 and diminish the dose rates (source term and radiation protection aspects), and

- limit and homogenise the deposit on the core and prevent axial offset anomaly (AOA) phenomena (fuel aspects).

Nevertheless, the positive impact of zinc injection on radiation field is still on debate at the international level.

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

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

- the curative mode: zinc is injected after SGR (or after the start-up of a new reactor).

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

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

Figure 11 to Figure 14 present Co-58 and Co-60 surface activities, for different sub-regions of the RCS, for each injection mode. Data related to units without (respectively with) zinc injection are in blue (respectively orange).
* Preventive mode (900 MWe series):

Figure 11 gives a general view of every Co-58 and Co-60 surface activities measurements in the RCS after oxygenation, given the number of cycles after SGR. This type of graphical representation can give the impression that units with zinc injection have a lower level of Co-58 surface contamination, and a medium to higher level of Co-60 contamination. Although, it might not be relevant to draw conclusions based on these figures only, since they do not allow to identify if one measurement (blue point or orange triangle) belongs to an upward or downward trend. Thus, they are completed by Figure 12, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to unit injecting zinc are displayed in colour (reactors B, C, D, E and F); those corresponding to units without zinc injection are in grey.

<figure><img src="image_17.png" /><figcaption>Figure 11: After 900 MWe SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection</figcaption></figure>

<figure><img src="image_18.png" /><figcaption>Figure 12: After 900 MWe SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)</figcaption></figure>

Co-58 mean surface contamination after SGR tends to decrease in both cases, with and without zinc injection. But the decrease appears to be sharper when units are injecting zinc. When the analysis focuses on the different parts of the RCS, this positive impact is significant in cross-over legs and SG (cold part) (see §6). In the hot legs, a similar effect of zinc injection is not detected.

Regarding Co-60 mean surface contamination in the RCS after SGR, first a downward trend can be observed for units without zinc injection, sometimes followed by a slow upward trend. In contrast, for units injecting zinc, an increase trend occurs quickly and lead to higher Co-60 surface contamination in comparison to units without zinc injection. When the analysis focuses on the different parts of the RCS, this effect of zinc injection appears mainly into the hot legs (see § 6).
* Preventive mode (900 MWe series):

Figure 11 gives a general view of every Co-58 and Co-60 surface activities measurements in the RCS after oxygenation, given the number of cycles after SGR. This type of graphical representation can give the impression that units with zinc injection have a lower level of Co-58 surface contamination, and a medium to higher level of Co-60 contamination. Although, it might not be relevant to draw conclusions based on these figures only, since they do not allow to identify if one measurement (blue point or orange triangle) belongs to an upward or downward trend. Thus, they are completed by Figure 12, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to unit injecting zinc are displayed in colour (reactors B, C, D, E and F); those corresponding to units without zinc injection are in grey.

<figure><img src="image_16.png" /><figcaption>Figure 11: After 900 MWe SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection</figcaption></figure>

<figure><img src="image_17.png" /><figcaption>Figure 12: After 900 MWe SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)</figcaption></figure>

Co-58 mean surface contamination after SGR tends to decrease in both cases, with and without zinc injection. But the decrease appears to be sharper when units are injecting zinc. When the analysis focuses on the different parts of the RCS, this positive impact is significant in cross-over legs and SG (cold part) (see §6). In the hot legs, a similar effect of zinc injection is not detected.

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

Figure 13 presents the general aspect of Co-58 and Co-60 surface activities given the number of cycles. As previously, these are completed by Figure 14, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to units injecting zinc are displayed in colour; those corresponding to units without zinc injection are in grey.

The beginning of zinc injection is indicated by red triangles (reminder: in curative mode, injections starts while the SG tubing surfaces are already passivated).

<figure><img src="image_19.png" /><figcaption>Figure 13: Before SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection</figcaption></figure>

<figure><img src="image_20.png" /><figcaption>Figure 14: Before SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)</figcaption></figure>

For both Co-58 and Co-60 deposited activities, the trend of the coloured curves after zinc injection does not seem to differ to the one without zinc injection. But we can note that the period of time after zinc is injected is too short to provide a robust analysis. Moreover, when zinc starts to be injected, surface activities are already quite low, and a modification of the trend is more difficult to detect.

## 4. Recontamination during the cold shutdown procedure

Regarding source term management, the shutdown procedure of a reactor is of great concern, the risk of out-of-flux surface recontamination by corrosion products being one of the sensitive points. The shutdown procedure is characterized by a progressive decrease of the temperature and the pH, and the evolution of the redox potential (reducing to oxidizing conditions). The forced oxygenation implemented on the French fleet leads to the dissolution of corrosion products and the occurrence of activated corrosion products peaks, especially in dissolved Co-58. It is now well known that this Co-58
* Curative mode:

Figure 13 presents the general aspect of Co-58 and Co-60 surface activities given the number of cycles. As previously, these are completed by Figure 14, which allows the comparison of the mean surface activity evolutions unit by unit. Measurements related to units injecting zinc are displayed in colour; those corresponding to units without zinc injection are in grey.

The beginning of zinc injection is indicated by red triangles (reminder: in curative mode, injections starts while the SG tubing surfaces are already passivated).

<figure><img src="image_18.png" /><figcaption>Figure 13: Before SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) after oxygenation with and without zinc injection</figcaption></figure>

<figure><img src="image_19.png" /><figcaption>Figure 14: Before SGR – Evolution of Co-58 (left) and Co-60 (right) surface activities in the RCS (HL, CL, COL, SG) by unit, with and without zinc injection (average values)</figcaption></figure>

For both Co-58 and Co-60 deposited activities, the trend of the coloured curves after zinc injection does not seem to differ to the one without zinc injection. But we can note that the period of time after zinc is injected is too short to provide a robust analysis. Moreover, when zinc starts to be injected, surface activities are already quite low, and a modification of the trend is more difficult to detect.

## 4. Recontamination during the cold shutdown procedure

Regarding source term management, the shutdown procedure of a reactor is of great concern, the risk of out-of-flux surface recontamination by corrosion products being one of the sensitive points. The shutdown procedure is characterized by a progressive decrease of the temperature and the pH, and the evolution of the redox potential (reducing to oxidizing conditions). The forced oxygenation implemented on the French fleet leads to the dissolution of corrosion products and the occurrence of activated corrosion products peaks, especially in dissolved Co-58. It is now well known that this Co-58
activity is due to the dissolution of metal nickel (and possibly of nickel oxide) in-core deposits, which leads to the dissolution of the Co-58 derived from the nickel sites in the crud ([11]). Effective purification means (flow rate, cationic and/or anionic resins with the adequate chemical conditioning...) are required in order to avoid a recontamination (redeposition by precipitation) of the primary (and auxiliary) surfaces by activated corrosion products. To avoid transferring this considerable activity from in-core surfaces to out-of-flux surfaces, it is also required to start the oxygenation phase at a maximum temperature of 80 °C, and to reduce the duration between the chemical dehydrogenation and the oxygenation strictly speaking (when partial oxygenation is implemented with a chemical dehydrogenation). It has been shown, for example in [12], that a recontamination of the primary circuit can occur if the oxygenation temperature of the primary coolant is higher than 80 °C.

It is not excluded that Co-58 recontamination is accompanied by a cobalt and nickel redeposition. In such a situation, these elements are transferred and activated in the core during the following cycle, creating an additional source of radioactive corrosion products.

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

## 4.1. Co-58 surface recontamination

Figure 15 and Figure 16 present the after to before ratios of Co-58 surface activities given the number of cycles, for the hot legs and the cold parts of the SG tube bundles respectively.

<figure><img src="image_21.png" /><figcaption>Figure 15: 900 MWe series - Evolution of Co-58 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

<figure><img src="image_22.png" /><figcaption>Figure 16: 1300 MWe series - Evolution of Co-58 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>
activity is due to the dissolution of metal nickel (and possibly of nickel oxide) in-core deposits, which leads to the dissolution of the Co-58 derived from the nickel sites in the crud ([11]). Effective purification means (flow rate, cationic and/or anionic resins with the adequate chemical conditioning...) are required in order to avoid a recontamination (redeposition by precipitation) of the primary (and auxiliary) surfaces by activated corrosion products. To avoid transferring this considerable activity from in-core surfaces to out-of-flux surfaces, it is also required to start the oxygenation phase at a maximum temperature of 80 °C, and to reduce the duration between the chemical dehydrogenation and the oxygenation strictly speaking (when partial oxygenation is implemented with a chemical dehydrogenation). It has been shown, for example in [12], that a recontamination of the primary circuit can occur if the oxygenation temperature of the primary coolant is higher than 80 °C.

It is not excluded that Co-58 recontamination is accompanied by a cobalt and nickel redeposition. In such a situation, these elements are transferred and activated in the core during the following cycle, creating an additional source of radioactive corrosion products.

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

## 4.1. Co-58 surface recontamination

Figure 15 and Figure 16 present the after to before ratios of Co-58 surface activities given the number of cycles, for the hot legs and the cold parts of the SG tube bundles respectively.

<figure><img src="image_20.png" /><figcaption>Figure 15: 900 MWe series - Evolution of Co-58 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

<figure><img src="image_21.png" /><figcaption>Figure 16: 1300 MWe series - Evolution of Co-58 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>
Co-58 recontamination occurs in the HL and SG tubing (cold part) for both series. Nevertheless, for the 900 MWe series, the proportion of recontamination events seems to diminish with time. For the 1300 MWe series, the recontamination of the surfaces during oxygenation appears to be quite frequent into the SG tubing (cold part) sub-region.

## 4.2. Co-60 surface recontamination

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

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

<figure><img src="image_23.png" /><figcaption>Figure 17: 900 MWe series - Evolution of Co-60 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

<figure><img src="image_24.png" /><figcaption>Figure 18: 1300 MWe series - Evolution of Co-60 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

# 5. Manufacturing processes of SG tubes made in 690TT alloy

This study, as many others, outlines the great impact of SGR on Co-58 and Co-60 surface contamination of the RCS. Several papers based on operating experience data (for example [3], [8], [9]) highlighted the prominent role of the surface finish of the SG tubes related to their manufacturing process (more than the tube alloy) in the Ni-58 and Co-59 releases and thus the contamination of the circuits.

In this §, the objective is to check whether the manufacturing processes of a similar alloy is likely to also have an effect on surface contamination. Here, we considered EMECC data related to the 900 MWe series after SGR. All new SGs are equipped with 690 alloy, as mentioned previously, but different manufacturing processes have been used.
Co-58 recontamination occurs in the HL and SG tubing (cold part) for both series. Nevertheless, for the 900 MWe series, the proportion of recontamination events seems to diminish with time. For the 1300 MWe series, the recontamination of the surfaces during oxygenation appears to be quite frequent into the SG tubing (cold part) sub-region.

## 4.2. Co-60 surface recontamination

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

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

<figure><img src="image_22.png" /></figure>

<figure><img src="image_23.png" /><figcaption>Figure 17: 900 MWe series - Evolution of Co-60 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

<figure><img src="image_24.png" /></figure>

<figure><img src="image_25.png" /><figcaption>Figure 18: 1300 MWe series - Evolution of Co-60 recontamination during the cold shutdown procedure in the HL (left) and SG tubing (cold part) (right)</figcaption></figure>

# 5. Manufacturing processes of SG tubes made in 690TT alloy

This study, as many others, outlines the great impact of SGR on Co-58 and Co-60 surface contamination of the RCS. Several papers based on operating experience data (for example [3], [8], [9]) highlighted the prominent role of the surface finish of the SG tubes related to their manufacturing process (more than the tube alloy) in the Ni-58 and Co-59 releases and thus the contamination of the circuits.

In this §, the objective is to check whether the manufacturing processes of a similar alloy is likely to also have an effect on surface contamination. Here, we considered EMECC data related to the 900 MWe series after SGR. All new SGs are equipped with 690 alloy, as mentioned previously, but different manufacturing processes have been used.
The impact of SG tube manufacturing processes and surface state have also been studied in papers using loop test experiments and characterization of both the internal surface of the tubes and the oxides formed (for example [13], [14], [15]).

As explained in [13], since the first use of 690 alloy for SG tubes in France, the manufacturing processes have been modified: the sandblasting with corundum particles has been suppressed, the annealing initially realized under NH₃ or H₂ is only performed under H₂ now, the carbon content has been reduced and the equivalent carbon content is taken into account for determining the annealing temperature, steps of pilgering and/or drawing have been optimized, cleanings have been introduced, cleaning baths have been modified etc.

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

## 5.1. Co-58 surface contamination

For each manufacturing process (“1” to “5”), Figure 19 and Figure 20 present the general aspect of Co-58 mean surface activities given the number of cycles after SGR.

<figure><img src="image_25.png" /><figcaption>Figure 19: 900 MWe series – Evolution of Co-58 surface activities in the HL (left) and CL&COL (right) given the manufacturing process</figcaption></figure>

<figure><img src="image_26.png" /><figcaption>Figure 20: 900 MWe series – Evolution of Co-58 surface activities in the SG, hot part (left) and cold part (right) given the manufacturing process</figcaption></figure>

According to these data, for every region of the RCS, all manufacturing processes share a common global trend (sharp decrease after a few cycles followed by a slower one until a plateau after around 10 cycles). After 10 cycles, Co-58 surface contamination levels are very close (except, maybe, for the hot legs, where a discrepancy remains). The most recent manufacturing process (n°5) seems to
The impact of SG tube manufacturing processes and surface state have also been studied in papers using loop test experiments and characterization of both the internal surface of the tubes and the oxides formed (for example [13], [14], [15]).

As explained in [13], since the first use of 690 alloy for SG tubes in France, the manufacturing processes have been modified: the sandblasting with corundum particles has been suppressed, the annealing initially realized under NH₃ or H₂ is only performed under H₂ now, the carbon content has been reduced and the equivalent carbon content is taken into account for determining the annealing temperature, steps of pilgering and/or drawing have been optimized, cleanings have been introduced, cleaning baths have been modified etc.

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

## 5.1. Co-58 surface contamination

For each manufacturing process (“1” to “5”), Figure 19 and Figure 20 present the general aspect of Co-58 mean surface activities given the number of cycles after SGR.

<figure><img src="image_26.png" /><figcaption>Figure 19: 900 MWe series – Evolution of Co-58 surface activities in the HL (left) and CL&COL (right) given the manufacturing process</figcaption></figure>

<figure><img src="image_27.png" /><figcaption>Figure 20: 900 MWe series – Evolution of Co-58 surface activities in the SG, hot part (left) and cold part (right) given the manufacturing process</figcaption></figure>

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

The oldest manufacturing process (n°1) presents high Co-58 surface contamination for the first cycles. In the paper [13], this manufacturing process was also associated to a slower decrease of the Co-58 spike height. Although, there is no direct relation between the out-of-flux Co-58 surface activities and the Co-58 oxygenation peak (which is linked to the source term deposited on the fuel), this is representative of a significant release of nickel from SG tubing.

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

## 5.2. Co-60 surface contamination

For each manufacturing process (“1” to “5”), Figure 21 and Figure 22 present the general aspect of Co-60 mean surface activities given the number of cycles after SGR.

<figure><img src="image_27.png" /><figcaption>Figure 21: 900 MWe series – Evolution of Co-60 surface activities in the HL (left) and CL&COL (right) given the manufacturing process</figcaption></figure>

<figure><img src="image_28.png" /><figcaption>Figure 22: 900 MWe series – Evolution of Co-60 surface activities in the SG, hot part (left) and cold part (right) given the manufacturing process</figcaption></figure>

These data show that, for every region of the RCS, all manufacturing processes present a common global trend (steady slow increase of the Co-60 surface contamination; the beginning of a decrease for the last cycles seems nevertheless to appear for the manufacturing process n°3).

The manufacturing process n°4 presents the best behavior, excepting at the beginning of the cycles in the hot legs, whereas the highest mean value on one cycle is reached for the manufacturing process n°5. However, it is impossible to conclude without a more comprehensive analysis, as many other factors influence the Co-60 surface activity evolution (stellites wear, cobalt impurities content in the SG tubes, contamination level of the pipes prior to SGR, unexpected material degradation, and maintenance operation, as mentioned previously).
present the best behavior (lower surface activity for the first cycles). Note that to strengthen this observation, the effect of the primary pipe Ni deposit level prior to SGR (“memory effect”) should be factored in.

The oldest manufacturing process (n°1) presents high Co-58 surface contamination for the first cycles. In the paper [13], this manufacturing process was also associated to a slower decrease of the Co-58 spike height. Although, there is no direct relation between the out-of-flux Co-58 surface activities and the Co-58 oxygenation peak (which is linked to the source term deposited on the fuel), this is representative of a significant release of nickel from SG tubing.

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

## 5.2. Co-60 surface contamination

For each manufacturing process (“1” to “5”), Figure 21 and Figure 22 present the general aspect of Co-60 mean surface activities given the number of cycles after SGR.

<figure><img src="image_28.png" /><figcaption>Figure 21: 900 MWe series – Evolution of Co-60 surface activities in the HL (left) and CL&COL (right) given the manufacturing process</figcaption></figure>

<figure><img src="image_29.png" /><figcaption>Figure 22: 900 MWe series – Evolution of Co-60 surface activities in the SG, hot part (left) and cold part (right) given the manufacturing process</figcaption></figure>

These data show that, for every region of the RCS, all manufacturing processes present a common global trend (steady slow increase of the Co-60 surface contamination; the beginning of a decrease for the last cycles seems nevertheless to appear for the manufacturing process n°3).

The manufacturing process n°4 presents the best behavior, excepting at the beginning of the cycles in the hot legs, whereas the highest mean value on one cycle is reached for the manufacturing process n°5. However, it is impossible to conclude without a more comprehensive analysis, as many other factors influence the Co-60 surface activity evolution (stellites wear, cobalt impurities content in the SG tubes, contamination level of the pipes prior to SGR, unexpected material degradation, and maintenance operation, as mentioned previously).
## 6. Discussion

The previous EMECC measurement analysis [3] outlined a “CP0 effect” among the reactors of the 900 MWe series. Indeed, before SGR, the primary circuit Co-60 contamination was significantly higher for CP0 units (the oldest French PWRs), compared to the CP1-CP2 ones, especially for the HL and CL (or cross-over legs) (less true for the SG tubing).

This discrepancy among the reactors of the 900 MWe series is reminded in the Figure 23 below.

<figure><img src="image_29.png" /><figcaption>Figure 23: 900 MWe – Comparison of Co-60 surface activities in the HL (left) and CL&COL (right) between the CP0 and CP1-CP2 sub-series before SGR (minimum/maximum and average values)</figcaption></figure>

<figure><img src="image_30.png" /><figcaption>Figure 24: 900 MWe – Comparison of Co-60 surface activities in the SG between the CP0 and CP1-CP2 sub-series after SGR (average values)</figcaption></figure>
## 6. Discussion

The previous EMECC measurement analysis [3] outlined a “CP0 effect” among the reactors of the 900 MWe series. Indeed, before SGR, the primary circuit Co-60 contamination was significantly higher for CP0 units (the oldest French PWRs), compared to the CP1-CP2 ones, especially for the HL and CL (or cross-over legs) (less true for the SG tubing).

This discrepancy among the reactors of the 900 MWe series is reminded in the Figure 23 below.

<figure><img src="image_30.png" /><figcaption>Figure 23: 900 MWe – Comparison of Co-60 surface activities in the HL (left) and CL&COL (right) between the CP0 and CP1-CP2 sub-series before SGR (minimum/maximum and average values)</figcaption></figure>

<figure><img src="image_31.png" /><figcaption>Figure 24: 900 MWe – Comparison of Co-60 surface activities in the SG between the CP0 and CP1-CP2 sub-series after SGR (average values)</figcaption></figure>
<figure><img src="image_31.png" /><figcaption>Figure 25: 900 MWe – Comparison of Co-60 surface activities between the CP0 and CP1-CP2 sub-series, in the HL (left) and CL&COL (right) (average values)</figcaption></figure>

After SGR, the updated data shows that the discrepancy between the 2 sub-series still remains on the primary coolant pipes but seems to disappear after around 10 cycles. In the SG tubing (see Figure 24), after SGR, the evolution of the Co-60 contamination is very similar when we compare the 2 sub-series (CP0 vs CP1-CP2).

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

In order to better characterize the different contributors to the Co-60 contamination, OSCAR calculations have been carried out and compared to EMECC measurements. The Figure 26 presents the evolution of the Co-60 contamination (for both measurements and calculation) before and after SGR of the reactor A, a 900 MWe series unit.

<figure><img src="image_32.png" /><figcaption>Figure 26: OSCAR calculation and EMECC measurements – Evolution of Co-60 surface contamination before and after SGR in the HL of the reactor A (900 MWe series)</figcaption></figure>

Based on the very good agreement between calculation and measurements, the main contributors of Co-60 contamination according to the OSCAR results have been distinguished (see Figure 26). The calculation show that three main categories of Co-60 sources can be identified:

- Stellites surfaces (wear and corrosion),

- SG tubing (releases of the Co-59 impurities into the tubing alloy),

- In core materials and stainless-steel piping.
<figure><img src="image_32.png" /><figcaption>Figure 25: 900 MWe – Comparison of Co-60 surface activities between the CP0 and CP1-CP2 sub-series, in the HL (left) and CL&COL (right) (average values)</figcaption></figure>

After SGR, the updated data shows that the discrepancy between the 2 sub-series still remains on the primary coolant pipes but seems to disappear after around 10 cycles. In the SG tubing (see Figure 24), after SGR, the evolution of the Co-60 contamination is very similar when we compare the 2 sub-series (CP0 vs CP1-CP2).

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

In order to better characterize the different contributors to the Co-60 contamination, OSCAR calculations have been carried out and compared to EMECC measurements. The Figure 26 presents the evolution of the Co-60 contamination (for both measurements and calculation) before and after SGR of the reactor A, a 900 MWe series unit.

<figure><img src="image_33.png" /><figcaption>Figure 26: OSCAR calculation and EMECC measurements – Evolution of Co-60 surface contamination before and after SGR in the HL of the reactor A (900 MWe series)</figcaption></figure>

Based on the very good agreement between calculation and measurements, the main contributors of Co-60 contamination according to the OSCAR results have been distinguished (see Figure 26). The calculation show that three main categories of Co-60 sources can be identified:

- Stellites surfaces (wear and corrosion),

- SG tubing (releases of the Co-59 impurities into the tubing alloy),

- In core materials and stainless-steel piping.
Before SGR, these three sources contribute to the Co-60 contamination approximatively equally, with a slight preponderance of the stellites source, especially the one associated to the CRDM and the CVCS pump. Note that the Co-60 contamination associated to "in core materials and stainless-steel piping" is far from being negligible: it comes from the contribution of, before their replacement, the Inconel spacer grids, which Co-59 impurity content were around 800 ppm. Just before SGR, the Co-60 contamination is almost stabilized. After SGR, the Co-60 increase rates due to stellites and in core materials and steel piping are a continuation of what they were just before SGR and are thus very low in comparison to their levels at the start-up of the reactor A (stellites wear and corrosion of materials are maximum at the beginning of life).

With the new SGs, the combination of a corrosion/release rate significantly lower than the previous one (replacement of 600 alloy by 690TT alloy and evolution of the manufacturing process) and a reduction of the Co-59 impurity content into the 690TT alloy tubing contributes to the very low increase of the Co-60 contamination. Nevertheless, we cannot conclude that SG are the main contributors to Co-60 surface contamination, as the "production" of each source of Co-59 evolves with time. As outlined previously in [8] and [9], the Co-60 build-up coming from the remaining sources is very low compared to the former Co-60 contamination in the primary coolant pipes. According to usual trends of the SG corrosion/release rate, a stable level of Co-60 contamination or even a decrease within a few cycles can be expected (due to the radioactive decay; providing there is no detrimental operating event).

## Zinc injection

The approach retained to analyze the possible effect of zinc injection on the surface contamination is to compare the evolution of the contamination of units with zinc injection to others without. In the case of the Co-60 contamination associated to the preventive mode injection, Figure 12 (right hand side) seems to show that the increase of the contamination after SGR is stronger and starts earlier for units B, C, D and E (which inject zinc) than for units without zinc injection. But data analysis can be tricky, and this preliminary conclusion needs to be strengthened by taking into account the radiological situation analysis of these units before SGR. Indeed, when considering the Co-60 deposited activities before SGR (and before zinc injection), some units may present lasting high contamination levels (resulting from stellites wear, operating events, maintenance operation...), temporarily lowered due to the SGR, which could also explain the increase trend after SGR. However, in the case of reactor B, the Co-60 surface activities evolution in the HL shows that high values have been reached (around 14 GBq/m²) after 12 cycles but a decrease trend in the following cycles leads to lower values of 4 to 5 GBq/m². Thus, it appears questionable to attribute the Co-60 surface contamination increase after SGR to the historical background of the plant before SGR.

Nevertheless, the finding that zinc injection could increase the Co-60 surface activities has to be moderated in the light of the following observation. Indeed, the global analysis focusing on average data (Figure 12, right hand side) coming from measurements in HL, CL, COL and SG may mask disparities in the evolution of the contamination depending on the sub-regions of the RCS considered. Figure 28 highlights this observation. Thus, in the SG, a positive effect of zinc injection regarding Co-60 surface contamination could begin to emerge, with a stabilization trend in the medium term of the deposited activities measured for reactors B, C, and D. Further data is needed to confirm this pattern. This increase of the Co-60 surface contamination could be linked to the antagonistic effect outlined in [16] regarding zinc injection: "when zinc is injected, Co dissolves from the inner oxide. A part of this surplus of Co is deposited on the fuel elements leading to an increase in the formation of Co-60. This activated deposit is eroded and the particles are deposited on the outer oxide of the out-of-core components, including the SG tubing, resulting in an increase in the Co-60 contamination of this outer oxide". Then, the effect of zinc on the corrosion, which lowers the release of Co-59 (passivation effect), could explain the stabilization of the surface contamination after the first cycles.
Before SGR, these three sources contribute to the Co-60 contamination approximatively equally, with a slight preponderance of the stellites source, especially the one associated to the CRDM and the CVCS pump. Note that the Co-60 contamination associated to "in core materials and stainless-steel piping" is far from being negligible: it comes from the contribution of, before their replacement, the Inconel spacer grids, which Co-59 impurity content were around 800 ppm. Just before SGR, the Co-60 contamination is almost stabilized. After SGR, the Co-60 increase rates due to stellites and in core materials and steel piping are a continuation of what they were just before SGR and are thus very low in comparison to their levels at the start-up of the reactor A (stellites wear and corrosion of materials are maximum at the beginning of life).

With the new SGs, the combination of a corrosion/release rate significantly lower than the previous one (replacement of 600 alloy by 690TT alloy and evolution of the manufacturing process) and a reduction of the Co-59 impurity content into the 690TT alloy tubing contributes to the very low increase of the Co-60 contamination. Nevertheless, we cannot conclude that SG are the main contributors to Co-60 surface contamination, as the "production" of each source of Co-59 evolves with time. As outlined previously in [8] and [9], the Co-60 build-up coming from the remaining sources is very low compared to the former Co-60 contamination in the primary coolant pipes. According to usual trends of the SG corrosion/release rate, a stable level of Co-60 contamination or even a decrease within a few cycles can be expected (due to the radioactive decay; providing there is no detrimental operating event).

## Zinc injection

The approach retained to analyze the possible effect of zinc injection on the surface contamination is to compare the evolution of the contamination of units with zinc injection to others without. In the case of the Co-60 contamination associated to the preventive mode injection, Figure 12 (right hand side) seems to show that the increase of the contamination after SGR is stronger and starts earlier for units B, C, D and E (which inject zinc) than for units without zinc injection. But data analysis can be tricky, and this preliminary conclusion needs to be strengthened by taking into account the radiological situation analysis of these units before SGR. Indeed, when considering the Co-60 deposited activities before SGR (and before zinc injection), some units may present lasting high contamination levels (resulting from stellites wear, operating events, maintenance operation...), temporarily lowered due to the SGR, which could also explain the increase trend after SGR. However, in the case of reactor B, the Co-60 surface activities evolution in the HL shows that high values have been reached (around 14 GBq/m²) after 12 cycles but a decrease trend in the following cycles leads to lower values of 4 to 5 GBq/m². Thus, it appears questionable to attribute the Co-60 surface contamination increase after SGR to the historical background of the plant before SGR.

Nevertheless, the finding that zinc injection could increase the Co-60 surface activities has to be moderated in the light of the following observation. Indeed, the global analysis focusing on average data (Figure 12, right hand side) coming from measurements in HL, CL, COL and SG may mask disparities in the evolution of the contamination depending on the sub-regions of the RCS considered. Figure 28 highlights this observation. Thus, in the SG, a positive effect of zinc injection regarding Co-60 surface contamination could begin to emerge, with a stabilization trend in the medium term of the deposited activities measured for reactors B, C, and D. Further data is needed to confirm this pattern. This increase of the Co-60 surface contamination could be linked to the antagonistic effect outlined in [16] regarding zinc injection: "when zinc is injected, Co dissolves from the inner oxide. A part of this surplus of Co is deposited on the fuel elements leading to an increase in the formation of Co-60. This activated deposit is eroded and the particles are deposited on the outer oxide of the out-of-core components, including the SG tubing, resulting in an increase in the Co-60 contamination of this outer oxide". Then, the effect of zinc on the corrosion, which lowers the release of Co-59 (passivation effect), could explain the stabilization of the surface contamination after the first cycles.
<figure><img src="image_33.png" /><figcaption>Figure 27: Reactor B – Evolution of Co-60 surface contamination in HL before SGR</figcaption></figure>

<figure><img src="image_34.png" /><figcaption>Figure 28: After SGR – Evolution of Co-60 surface activities in the HL (left) and SG (right) with and without zinc injection</figcaption></figure>

Note that in the case of Co-58 surface contamination after SGR associated to the preventive mode injection, the decrease trend is quite homogenous in the RCS, being observed in every sub-regions where measurements are performed. The discrepancy between units injecting zinc and units without zinc injection is nevertheless more marked into the SG sub-region (see Figure 29 below).

<figure><img src="image_35.png" /><figcaption>Figure 29: After SGR – Evolution of Co-58 surface activities in the SG with and without zinc injection</figcaption></figure>

To complete, we can mention here that reactors B, C, D and E are equipped with SG tubes manufactured with the process n°5 (the most recent one; see previously): although some reactors without zinc injection are also equipped with such SG tubes, given the observation presented in §5.1, this parameter may interfere in the analysis.
<figure><img src="image_34.png" /><figcaption>Figure 27: Reactor B – Evolution of Co-60 surface contamination in HL before SGR</figcaption></figure>

<figure><img src="image_35.png" /><figcaption>Figure 28: After SGR – Evolution of Co-60 surface activities in the HL (left) and SG (right) with and without zinc injection</figcaption></figure>

Note that in the case of Co-58 surface contamination after SGR associated to the preventive mode injection, the decrease trend is quite homogenous in the RCS, being observed in every sub-regions where measurements are performed. The discrepancy between units injecting zinc and units without zinc injection is nevertheless more marked into the SG sub-region (see Figure 29 below).

<figure><img src="image_36.png" /><figcaption>Figure 29: After SGR – Evolution of Co-58 surface activities in the SG with and without zinc injection</figcaption></figure>

To complete, we can mention here that reactors B, C, D and E are equipped with SG tubes manufactured with the process n°5 (the most recent one; see previously): although some reactors without zinc injection are also equipped with such SG tubes, given the observation presented in §5.1, this parameter may interfere in the analysis.
## 7. Conclusion

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

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

- For Co-58:

* The surface contamination values and the data scatter decrease with time as well as with the evolution of the reactor design (the most recent reactors are less contaminated than the older, for a same cycle),

* In the 900 MWe series, SGR allows to further lower the contamination levels. After a few cycles (the number of which depends on the RCS sub-region), every part of the primary circuit reaches a surface activity below the one prior to SGR.

- For Co-60:

* The positive effect expected of the newest reactor design is not so well marked, except in the SG sub-regions,

* After SGR, the updated data for the 900 MWe confirm the smooth increase trend of the contamination in the SG but do not confirm in the longer term the decrease trend initially observed in the HL and CL.

* After SGR, the analysis does confirm the "CP0 effect" outlined before SGR for the 900 MWe series but only for the mid-term: in the longer term, on the contrary, CP1-CP2 contamination levels tend to be higher than the CP0 ones and to exceed the values before SGR. As only a few data are available for the long-term period (> 15 cycles after SGR), it seems good to remain cautious in the interpretation of this observation.

### Regarding the impact of zinc injection:

- When zinc is injected prior to SGR (or at the very beginning of the cycle following the SGR), a favorable impact on Co-58 deposited activities is likely to be expected in the different parts of the RCS,

- To be relevant, such an analysis needs:

* To follow the evolution of the contamination unit by unit,

* To take into account the contamination historicity of each unit and especially the contamination levels prior to SGR,

* To distinguish the data according to the location of the measurements (e.g.: HL, CL, COL, SG bundle...).

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

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

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

## References

[1] R. Eimecke, S. Anthoni – « Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC) », 7th Int. Conf. On Radiation Shielding, Bournemouth (UK), 1988
[2] F. Dacquait and al., "Ag-110m behaviour in PWRs: Lessons learnt from the EMECC campaigns", 20th NPC International Conference Brighton, United Kingdom - October 2-7, 2016, Paper Number: 81

[3] F. Dacquait, L. Guinard, C. Andrieu, JL. 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, 11-14/10/04, San Francisco (USA), Session 2, paper 4, EPRI, Palo Alto (2005)

[4] F. Dacquait, J. Francescatto, G. Galassi, F. Broutin, M. Gherrab, D. You, T. Jobert and N. Engler, 2018, "The OSCAR V1.4 code: a new dissolution/precipitation model to better simulate the corrosion product transfer in nuclear cooling systems", International Conference on Water Chemistry in Nuclear Reactor Systems, NPC, 2018, San Francisco

[5] S. Anthoni, P. Ridoux, O. Menet and C. Weber, "Effects of pH of Primary Coolant on PWR Contamination", Proceedings of the International Conference of Water Chemistry of Nuclear Reactor systems 6, Bournemouth 12-15 October 1992, Vol. 2, p. 9-15, BNES, London (1992)

[6] JL. Bretelle, L. Guinard, C. Andrieu and F. Dacquait, "EDF Strategy for the PWR Primary Coolant pH", NPC, 2004, San Francisco

[7] T. Jobert and al., "Study of the impact of three B/Li coordinated chemistries on the cobalt 58 surface contamination of a French PWR using the OSCAR code v1.4", Proceedings of ICONE-27, 27th International Conference on Nuclear Engineering, May 19-24, 2019, Ibaraki, Japan (paper n°2236)

[8] F. Dacquait and al., "Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products", NPC, 2000

[9] F. Dacquait and al., "The impact of steam generator replacement on PWR primary system contamination", Proceedings of the Nuclear Plant Chemistry Conference 2010, 03-07/10/10, Quebec City (Canada), paper 5.04

[10] B. Phlippot and al., "15 year experience of zinc injection at EDF", Proceedings of the 21st International Conference on Water Chemistry in Nuclear Reactor Systems San Francisco, CA, United States - September 9-14, 2018

[11] F. Dacquait and al., "Simulation of corrosion product transfer during PWR cold shutdown using the OSCAR code", this conference

[12] F. Dacquait et al., "Corrosion product transfer in French PWRs during shutdown", Proceedings of the International conference on water chemistry in nuclear reactors systems, 22-26/04/02, Avignon (France), CDROM, paper 113, 2002

[13] F. Carrette and al., "Impact of the Surface State of Steam Generator Tubes on the Release of Corrosion Products in Pressurized Water Reactors", NPC, 2006, Jeju Island, Korea

[14] F. Carrette and al., "Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors", NPC, 2012, Paris, France

[15] M. Clauzel and N. Engler, "Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant", NPC, 2010, October 3-7, Quebec City (Canada)

[16] M. Bultot and al., « Zinc effect on the primary circuit contamination of a Belgian PWR using the OSCAR V1.3 code », Proceedings of the Nuclear Plant Chemistry Conference 2016, Brighton (UK), paper 80
[2] F. Dacquait and al., "Ag-110m behaviour in PWRs: Lessons learnt from the EMECC campaigns", 20th NPC International Conference Brighton, United Kingdom - October 2-7, 2016, Paper Number: 81

[3] F. Dacquait, L. Guinard, C. Andrieu, JL. 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, 11-14/10/04, San Francisco (USA), Session 2, paper 4, EPRI, Palo Alto (2005)

[4] F. Dacquait, J. Francescatto, G. Galassi, F. Broutin, M. Gherrab, D. You, T. Jobert and N. Engler, 2018, "The OSCAR V1.4 code: a new dissolution/precipitation model to better simulate the corrosion product transfer in nuclear cooling systems", International Conference on Water Chemistry in Nuclear Reactor Systems, NPC, 2018, San Francisco

[5] S. Anthoni, P. Ridoux, O. Menet and C. Weber, "Effects of pH of Primary Coolant on PWR Contamination", Proceedings of the International Conference of Water Chemistry of Nuclear Reactor systems 6, Bournemouth 12-15 October 1992, Vol. 2, p. 9-15, BNES, London (1992)

[6] JL. Bretelle, L. Guinard, C. Andrieu and F. Dacquait, "EDF Strategy for the PWR Primary Coolant pH", NPC, 2004, San Francisco

[7] T. Jobert and al., "Study of the impact of three B/Li coordinated chemistries on the cobalt 58 surface contamination of a French PWR using the OSCAR code v1.4", Proceedings of ICONE-27, 27th International Conference on Nuclear Engineering, May 19-24, 2019, Ibaraki, Japan (paper n°2236)

[8] F. Dacquait and al., "Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products", NPC, 2000

[9] F. Dacquait and al., "The impact of steam generator replacement on PWR primary system contamination", Proceedings of the Nuclear Plant Chemistry Conference 2010, 03-07/10/10, Quebec City (Canada), paper 5.04

[10] B. Phlippot and al., "15 year experience of zinc injection at EDF", Proceedings of the 21st International Conference on Water Chemistry in Nuclear Reactor Systems San Francisco, CA, United States - September 9-14, 2018

[11] F. Dacquait and al., "Simulation of corrosion product transfer during PWR cold shutdown using the OSCAR code", this conference

[12] F. Dacquait et al., "Corrosion product transfer in French PWRs during shutdown", Proceedings of the International conference on water chemistry in nuclear reactors systems, 22-26/04/02, Avignon (France), CDROM, paper 113, 2002

[13] F. Carrette and al., "Impact of the Surface State of Steam Generator Tubes on the Release of Corrosion Products in Pressurized Water Reactors", NPC, 2006, Jeju Island, Korea

[14] F. Carrette and al., "Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors", NPC, 2012, Paris, France

[15] M. Clauzel and N. Engler, "Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant", NPC, 2010, October 3-7, Quebec City (Canada)

[16] M. Bultot and al., « Zinc effect on the primary circuit contamination of a Belgian PWR using the OSCAR V1.3 code", Proceedings of the Nuclear Plant Chemistry Conference 2016, Brighton (UK), paper 80