# Impact of Steam Generator primary circuit contamination products on PWR primary circuit contamination by corrosion products.

Frédéric Dacquait, Serge Anthoni, Philippe Ridoux, Christian Brun, Nathalie Engler

► To cite this version:

Frédéric Dacquait, Serge Anthoni, Philippe Ridoux, Christian Brun, Nathalie Engler. Impact of Steam Generator primary circuit contamination products on PWR primary circuit contamination by corrosion products. International conference on water chemistry of nuclear reactor systems 8, Oct 2000, Buournemouth, United Kingdom. pp.53-58. cea-04464354

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Submitted on 18 Feb 2024

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# Impact of Steam Generator primary circuit contamination products on PWR primary circuit contamination by corrosion products.

Frédéric Dacquait, Serge Anthoni, Philippe Ridoux, Christian Brun, Nathalie Engler

► To cite this version:

Frédéric Dacquait, Serge Anthoni, Philippe Ridoux, Christian Brun, Nathalie Engler. Impact of Steam Generator primary circuit contamination products on PWR primary circuit contamination by corrosion products. International conference on water chemistry of nuclear reactor systems 8, Oct 2000, Buournemouth, United Kingdom. pp.53-58. cea-04464354

HAL Id: cea-04464354

https://cea.hal.science/cea-04464354v1

Submitted on 18 Feb 2024

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

L'archive ouverte pluridisciplinaire HAL, est
destinée au dépôt et à la diffusion de documents
scientifiques de niveau recherche, publiés ou non,
émanant des établissements d'enseignement et de
recherche français ou étrangers, des laboratoires
publics ou privés.

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

Distributed under a Creative Commons Attribution 4.0 International License
# Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products

F. Dacquait and S. Anthoni, CEA, Centre d'Etudes de Cadarache, France

P. Ridoux, EDF, SEPTEN, France

C. Brun and N. Engler, Framatome, France

This paper analyses the impact of Steam Generator Replacement (SGR) on PWR primary circuit contamination by corrosion products. It presents a comparison of deposited and released activities between three new SG tube materials: alloys 600TT, 690TT and 800. This comparison allows to describe the typical evolutions of $^{58}$Co and $^{60}$Co contamination due to SGR. The discussion highlights the importance of corrosion product deposits on primary surfaces before SGR and the manufacturing process of SG tubes.

## Introduction

During reactor maintenance, for most PWRs, the collective dose is mainly due to $^{58}$Co and $^{60}$Co. As shown in Figure 1, $^{58}$Co is predominant in dose rate for the first reactor cycles. Progressively, its contribution decreases giving way to $^{60}$Co, which accumulates because of its radioactive half-life. Nevertheless for some reactors (circles in Figure 1), after several cycles, the contribution of $^{58}$Co may reach the level of the first cycles.

<figure><img src="image_2.png" /><figcaption>Figure 1: Ratio $^{58}$Co/$^{60}$Co for deposits</figcaption></figure>

This paper will show that this increase is the result of Steam Generator Replacement (SGR).

Because of Inconel 600 Mill-Annealed tube corrosion problems (intergranular attack, stress corrosion cracking, ...), PWR steam generators have to be replaced. The new tube material can be Thermally-Treated Inconel 600 or 690, and Incoloy 800.

Regarding contamination, Inconel heat treatment should reduce the release [1], alloy 690 high Chromium content (about 29% Cr, 59% Ni, 10% Fe) is meant to reduce generalized corrosion and alloy 800 low Nickel content (about 33% Ni, 21% Cr, 44% Fe) should reduce the $^{58}$Co source.

The impact of these three tube materials on contamination is compared and typical evolutions of $^{58}$Co and $^{60}$Co contamination due to SGR are described and analysed.

## Deposited activities

To analyse SGR impact on contamination, deposited activities on out-of-flux primary surfaces have been measured, before and after SGR, for 13 PWRs in France and abroad:

* 10 SGRs with alloy 690TT tubes which have undergone different manufacturing processes,

* 1 SGR with alloy 600TT tubes,

* 2 SGRs with alloy 800 tubes.

These PWRs have operated with coolant $pH_{300°C}$ of 7.0 or 7.2 and at base load or load follow.

The deposited activities have been measured on the hot legs, the crossover legs and the SG tube bundles by gamma spectrometry with the EMECC system [2].

Because of their important contribution to dose rate in the PWRs concerned, only the two main cobalt isotopes, $^{58}$Co and $^{60}$Co, are taken into account. Let us recall that $^{58}$Co (radioactive half-life: 71 days) comes from $^{58}$Ni (68% of natural Nickel) and $^{60}$Co (radioactive half-life: 5.3 years) comes from $^{59}$Co (100% of natural Cobalt).

### $^{58}$Co contamination

Figure 2 and Figure 3 show the variation of the $^{58}$Co deposited activities on the primary coolant pipes (hot legs and crossover legs) for the three types of SGR. The reactor
# Impact of Steam Generator Replacement on PWR primary circuit contamination by radioactive corrosion products

F. Dacquait and S. Anthoni, CEA, Centre d'Etudes de Cadarache, France

P. Ridoux, EDF, SEPTEN, France

C. Brun and N. Engler, Framatome, France

This paper analyses the impact of Steam Generator Replacement (SGR) on PWR primary circuit contamination by corrosion products. It presents a comparison of deposited and released activities between three new SG tube materials: alloys 600TT, 690TT and 800. This comparison allows to describe the typical evolutions of $^{58}$Co and $^{60}$Co contamination due to SGR. The discussion highlights the importance of corrosion product deposits on primary surfaces before SGR and the manufacturing process of SG tubes.

## Introduction

During reactor maintenance, for most PWRs, the collective dose is mainly due to $^{58}$Co and $^{60}$Co. As shown in Figure 1, $^{58}$Co is predominant in dose rate for the first reactor cycles. Progressively, its contribution decreases giving way to $^{60}$Co, which accumulates because of its radioactive half-life. Nevertheless for some reactors (circles in Figure 1), after several cycles, the contribution of $^{58}$Co may reach the level of the first cycles.

<figure><img src="image_2.png" /><figcaption>Figure 1 : Ratio $^{58}$Co/$^{60}$Co for deposits</figcaption></figure>

This paper will show that this increase is the result of Steam Generator Replacement (SGR).

Because of Inconel 600 Mill-Annealed tube corrosion problems (intergranular attack, stress corrosion cracking, ...), PWR steam generators have to be replaced. The new tube material can be Thermally-Treated Inconel 600 or 690, and Incoloy 800.

Regarding contamination, Inconel heat treatment should reduce the release [1], alloy 690 high Chromium content (about 29% Cr, 59% Ni, 10% Fe) is meant to reduce generalized corrosion and alloy 800 low Nickel content (about 33% Ni, 21% Cr, 44% Fe) should reduce the $^{58}$Co source.

The impact of these three tube materials on contamination is compared and typical evolutions of $^{58}$Co and $^{60}$Co contamination due to SGR are described and analysed.

## Deposited activities

To analyse SGR impact on contamination, deposited activities on out-of-flux primary surfaces have been measured, before and after SGR, for 13 PWRs in France and abroad :

* 10 SGRs with alloy 690TT tubes which have undergone different manufacturing processes,

* 1 SGR with alloy 600TT tubes,

* 2 SGRs with alloy 800 tubes.

These PWRs have operated with coolant $pH_{300°C}$ of 7.0 or 7.2 and at base load or load follow.

The deposited activities have been measured on the hot legs, the crossover legs and the SG tube bundles by gamma spectrometry with the EMECC system [2].

Because of their important contribution to dose rate in the PWRs concerned, only the two main cobalt isotopes, $^{58}$Co and $^{60}$Co, are taken into account. Let us recall that $^{58}$Co (radioactive half-life : 71 days) comes from $^{58}$Ni (68% of natural Nickel) and $^{60}$Co (radioactive half-life : 5.3 years) comes from $^{59}$Co (100% of natural Cobalt).

### $^{58}$Co contamination

Figure 2 and Figure 3 show the variation of the $^{58}$Co deposited activities on the primary coolant pipes (hot legs and crossover legs) for the three types of SGR. The reactor
«A» is representative of PWRs that have undergone a SGR with alloy 690TT.

Whatever the new tube material is (alloys 600TT, 690TT or 800), the ⁵⁸Co deposited activity on the primary piping increases for one or two cycles after SGR, and then decreases. Figure 4 shows the different hot leg contamination rises in PWRs that had a SGR with alloy 690TT.

<figure><img src="image_3.png" /><figcaption>Figure 2 : ⁵⁸Co deposited activities on the hot legs</figcaption></figure>

<figure><img src="image_4.png" /><figcaption>Figure 3 : ⁵⁸Co deposited activities on the crossover legs</figcaption></figure>

<figure><img src="image_5.png" /><figcaption>Figure 4 : ⁵⁸Co deposited activities on the hot legs</figcaption></figure>

At the first cycle, after SGR with Inconel, the ⁵⁸Co deposited activity on the SG tubes is generally lower than before SGR (Figure 5). Then, it increases for one or two cycles and decreases. For the reactor « C » (SGR with alloy 800), after a rise at the first cycle after SGR, the ⁵⁸Co deposited activity on the SG tubes slightly decreases on the cold side but highly drops on the hot side. Thus, at the sixth cycle after SGR, the cold side is six times more contaminated than the hot side. After a SGR with alloy 800, the oxide structure must be such as its solubility is very sensitive to temperature. It should be observed that there is practically no contamination difference between the two sides of Inconel SG tubes. If there is any difference, it is lower and the hot side is generally more contaminated.

<figure><img src="image_6.png" /><figcaption>Figure 5 : ⁵⁸Co deposited activities on the SG tubes</figcaption></figure>
«A» is representative of PWRs that have undergone a SGR with alloy 690TT.

Whatever the new tube material is (alloys 600TT, 690TT or 800), the ⁵⁸Co deposited activity on the primary piping increases for one or two cycles after SGR, and then decreases. Figure 4 shows the different hot leg contamination rises in PWRs that had a SGR with alloy 690TT.

<figure><img src="image_3.png" /><figcaption>Figure 2 : ⁵⁸Co deposited activities on the hot legs</figcaption></figure>

<figure><img src="image_4.png" /><figcaption>Figure 3 : ⁵⁸Co deposited activities on the crossover legs</figcaption></figure>

<figure><img src="image_5.png" /><figcaption>Figure 4 : ⁵⁸Co deposited activities on the hot legs</figcaption></figure>

At the first cycle, after SGR with Inconel, the ⁵⁸Co deposited activity on the SG tubes is generally lower than before SGR (Figure 5). Then, it increases for one or two cycles and decreases. For the reactor « C » (SGR with alloy 800), after a rise at the first cycle after SGR, the ⁵⁸Co deposited activity on the SG tubes slightly decreases on the cold side but highly drops on the hot side. Thus, at the sixth cycle after SGR, the cold side is six times more contaminated than the hot side. After a SGR with alloy 800, the oxide structure must be such as its solubility is very sensitive to temperature. It should be observed that there is practically no contamination difference between the two sides of Inconel SG tubes. If there is any difference, it is lower and the hot side is generally more contaminated.

<figure><img src="image_6.png" /><figcaption>Figure 5 : ⁵⁸Co deposited activities on the SG tubes</figcaption></figure>
<figure><img src="image_7.png" /><figcaption>Figure 6 : $^{58}$Co deposited activities on the SG tubes</figcaption></figure>

As seen in Figure 6 (reactor « G »), after a SGR with alloy 690TT, the $^{58}$Co deposited activity on the tube bundles can reach a low level comparable to the one measured on the hot side of alloy 800 SG tubes. Thus, for a same SG tube material (alloy 690TT), the tube bundle contamination can be really different. The manufacturing process of the SG tubes could explain these differences. Indeed, e.g. for the reactor « A » before SGR (Figure 5), the alloy 600MA SG tubes manufactured by « Y » are about four times more contaminated than those manufactured by « X ». It is important to recall that these SG tubes have undergone the same operating conditions (fluid chemistry, thermohydraulic, ...).

The impact of SGR on $^{58}$Co contamination distribution in the primary circuit is illustrated in Figure 7. This figure shows that, because of SGR, there is a higher $^{58}$Co deposited activity increase on the primary coolant pipes than on the tube bundles.

<figure><img src="image_8.png" /><figcaption>Figure 7 : Ratio primary pipes/SGs for $^{58}$Co</figcaption></figure>

$^{60}$Co contamination

The $^{60}$Co deposited activities on the primary system are represented from Figure 8 to Figure 10 for the three types of SGR.

Whatever the SG tube material is, since SGR the $^{60}$Co contamination has decreased on the primary coolant piping (Figure 8 and Figure 9). For the reactor « A », the $^{60}$Co deposited activity tends to stabilize after 4 cycles of reduction. For the crossover legs of the reactor « C », the decrease practically follows the radioactive decay.

The $^{60}$Co contamination rises on the new tube bundles for the three types of tube material (Figure 10); except on the SG hot side of the reactor « C » (SGR with alloy 800), where the $^{60}$Co deposited activity decreases after an increase. As well as for the $^{58}$Co contamination, the SG hot side is less contaminated by $^{60}$Co than the cold side (three times less).

<figure><img src="image_9.png" /><figcaption>Figure 8 : $^{60}$Co deposited activities on the hot legs</figcaption></figure>

<figure><img src="image_10.png" /><figcaption>Figure 9 : $^{60}$Co deposited activities on the crossover legs</figcaption></figure>
<figure><img src="image_7.png" /><figcaption>Figure 6 : $^{58}$Co deposited activities on the SG tubes</figcaption></figure>

As seen in Figure 6 (reactor « G »), after a SGR with alloy 690TT, the $^{58}$Co deposited activity on the tube bundles can reach a low level comparable to the one measured on the hot side of alloy 800 SG tubes. Thus, for a same SG tube material (alloy 690TT), the tube bundle contamination can be really different. The manufacturing process of the SG tubes could explain these differences. Indeed, e.g. for the reactor « A » before SGR (Figure 5), the alloy 600MA SG tubes manufactured by « Y » are about four times more contaminated than those manufactured by « X ». It is important to recall that these SG tubes have undergone the same operating conditions (fluid chemistry, thermohydraulic, ...).

The impact of SGR on $^{58}$Co contamination distribution in the primary circuit is illustrated in Figure 7. This figure shows that, because of SGR, there is a higher $^{58}$Co deposited activity increase on the primary coolant pipes than on the tube bundles.

<figure><img src="image_8.png" /><figcaption>Figure 7 : Ratio primary pipes/SGs for $^{58}$Co</figcaption></figure>

$^{60}$Co contamination

The $^{60}$Co deposited activities on the primary system are represented from Figure 8 to Figure 10 for the three types of SGR.

Whatever the SG tube material is, since SGR the $^{60}$Co contamination has decreased on the primary coolant piping (Figure 8 and Figure 9). For the reactor « A », the $^{60}$Co deposited activity tends to stabilize after 4 cycles of reduction. For the crossover legs of the reactor « C », the decrease practically follows the radioactive decay.

The $^{60}$Co contamination rises on the new tube bundles for the three types of tube material (Figure 10); except on the SG hot side of the reactor « C » (SGR with alloy 800), where the $^{60}$Co deposited activity decreases after an increase. As well as for the $^{58}$Co contamination, the SG hot side is less contaminated by $^{60}$Co than the cold side (three times less).

<figure><img src="image_9.png" /><figcaption>Figure 8 : $^{60}$Co deposited activities on the hot legs</figcaption></figure>

<figure><img src="image_10.png" /><figcaption>Figure 9 : $^{60}$Co deposited activities on the crossover legs</figcaption></figure>
<figure><img src="image_11.png" /><figcaption>Figure 10 : $^{60}$Co deposited activity on the SG tubes</figcaption></figure>

Figure 11 gives the variation of the $^{60}$Co contamination distribution in the primary circuit because of SGR. The increase of the ratio « primary coolant pipes/SG tubes » is due to the very low level of the $^{60}$Co deposited activity on the new SG tubes.

<figure><img src="image_12.png" /><figcaption>Figure 11 : Ratio primary pipes/SGs for $^{60}$Co</figcaption></figure>

### $^{58}$Co peak

Figure 12 shows the SGR impact on the $^{58}$Co concentration peak during refuelling shutdown. Let us recall that the $^{58}$Co peak comes from the dissolution of deposited Nickel (Ni⁰ or NiO) on the fuel rods because of primary system oxygenation [3].

The $^{58}$Co peak variation due to a SGR is similar to the $^{58}$Co deposited activity on the primary coolant piping : the $^{58}$Co peak rises generally for one or two cycles after SGR, and then decreases.

The lower peak of the reactor « C » is not due to the SGR with alloy 800 tubes (the peak was already low before the SGR even when the SG tube material was alloy 600MA) but because of the shutdown procedure. Indeed, the total $^{58}$Co volume activity released during shutdowns

of the reactor « C » is equivalent to the one of a reactor with alloy 690TT tubes (about 30,000 GBq).

It should be noted that the reactor « H » (alloy 690TT SG tubes) presents very low peaks and total released activities. It is not the result of the shutdown procedure but it is due to low nickel deposits on the fuel rods very likely thanks to lower temperatures of the fuel rod claddings.

<figure><img src="image_13.png" /><figcaption>Figure 12 : $^{58}$Co concentration peak during shutdown</figcaption></figure>

### Dose rates

Figure 13 shows the dose rate around the primary coolant pipes for PWRs having undergone a SGR with Inconel. The dose rate increases generally for one or two cycles and then decreases because of the $^{58}$Co deposited activity variation.

The dose rate at the primary coolant pipes reflecting the combined effect of the radionuclides and especially the ambient dose rate [4], it can be stable after a SGR even if $^{58}$Co deposited activity on the primary coolant pipes rises (e.g. reactor « F »).

<figure><img src="image_14.png" /><figcaption>Figure 13 : Dose rate at the reactor coolant pipes</figcaption></figure>
<figure><img src="image_11.png" /><figcaption>Figure 10 : $^{60}$Co deposited activity on the SG tubes</figcaption></figure>

Figure 11 gives the variation of the $^{60}$Co contamination distribution in the primary circuit because of SGR. The increase of the ratio « primary coolant pipes/SG tubes » is due to the very low level of the $^{60}$Co deposited activity on the new SG tubes.

<figure><img src="image_12.png" /><figcaption>Figure 11 : Ratio primary pipes/SGs for $^{60}$Co</figcaption></figure>

### $^{58}$Co peak

Figure 12 shows the SGR impact on the $^{58}$Co concentration peak during refuelling shutdown. Let us recall that the $^{58}$Co peak comes from the dissolution of deposited Nickel (Ni⁰ or NiO) on the fuel rods because of primary system oxygenation [3].

The $^{58}$Co peak variation due to a SGR is similar to the $^{58}$Co deposited activity on the primary coolant piping : the $^{58}$Co peak rises generally for one or two cycles after SGR, and then decreases.

The lower peak of the reactor « C » is not due to the SGR with alloy 800 tubes (the peak was already low before the SGR even when the SG tube material was alloy 600MA) but because of the shutdown procedure. Indeed, the total $^{58}$Co volume activity released during shutdowns

of the reactor « C » is equivalent to the one of a reactor with alloy 690TT tubes (about 30,000 GBq).

It should be noted that the reactor « H » (alloy 690TT SG tubes) presents very low peaks and total released activities. It is not the result of the shutdown procedure but it is due to low nickel deposits on the fuel rods very likely thanks to lower temperatures of the fuel rod claddings.

<figure><img src="image_13.png" /><figcaption>Figure 12 : $^{58}$Co concentration peak during shutdown</figcaption></figure>

### Dose rates

Figure 13 shows the dose rate around the primary coolant pipes for PWRs having undergone a SGR with Inconel. The dose rate increases generally for one or two cycles and then decreases because of the $^{58}$Co deposited activity variation.

The dose rate at the primary coolant pipes reflecting the combined effect of the radionuclides and especially the ambient dose rate [4], it can be stable after a SGR even if $^{58}$Co deposited activity on the primary coolant pipes rises (e.g. reactor « F »).

<figure><img src="image_14.png" /><figcaption>Figure 13 : Dose rate at the reactor coolant pipes</figcaption></figure>
Thanks to the low <sup>60</sup>Co contamination on the SG tubes, the dose rate around the SG tube bundles is low for the first cycle after SGR (● Figure 14). It increases with the radioactive cobalt contamination and then decreases with <sup>58</sup>Co.

<figure><img src="image_15.png" /><figcaption>Figure 14 : Dose rate at the SG tube bundles</figcaption></figure>

### <sup>58</sup>Co and <sup>60</sup>Co Contribution

Only <sup>58</sup>Co and <sup>60</sup>Co are taken into account to evaluate the radionuclide contribution, even if they represent about 90% of the dose rate for the reactors concerned.

Figure 15 and Figure 16 show the <sup>58</sup>Co contribution to the dose rate in the vicinity of the crossover legs and SGs, respectively.

Just before SGR, <sup>60</sup>Co contributes to 70%-80% of the dose rate at primary system. Because of SGR, the <sup>58</sup>Co contribution increases and generally becomes the major one. The <sup>58</sup>Co contribution reaches the levels preceding SGR with alloy 800 much faster than SGR with Inconel.

<figure><img src="image_16.png" /><figcaption>Figure 15 : <sup>58</sup>Co contribution to the dose rate at the crossover legs</figcaption></figure>

<figure><img src="image_17.png" /><figcaption>Figure 16 : <sup>58</sup>Co contribution to the dose rate at the SG</figcaption></figure>

## Discussion

### <sup>58</sup>Co contamination

Because of SGR, Nickel release increases in the primary system up to three cycles. This rise is shown in an increase of the <sup>58</sup>Co deposited activity on the primary coolant loops and an increase of <sup>58</sup>Co activity released from fuel rods during shutdown with oxygenation. Then, the <sup>58</sup>Co contamination decreases for several cycles and generally reaches lower levels than before SGR.

These levels do not seem to mainly depend on alloy composition :

* a SGR with alloy 600TT can lead to a <sup>58</sup>Co contamination similar to a SGR with alloy 690TT,

* the <sup>58</sup>Co deposited activity on the alloy 690TT SG tubes can be comparable to the one on the alloy 800 SG tubes,

* the decrease of <sup>58</sup>Co contamination, after a SGR with alloy 800, is less than expected ; except on the hot side of primary coolant loops. This relative high <sup>58</sup>Co contamination shows that there is still a large source of Nickel after a SGR with alloy 800.

After a SGR, the <sup>58</sup>Co variations can be due to :

* a release and a trapping of SG tubes depending on the manufacturing process of new SG tubes [1] [5],

* a higher release of SG tubes during formation of the inner protective oxide layer,

* a « memory effect » due to Nickel deposits in the primary system before SGR, this effect decreasing with this Nickel reserve.

In any case, the contamination levels reached after SGR are not due to reactor coolant chemistry as long as pH<sub>300°C</sub> lies between 7.0 and 7.4. As there is no correlation with reactor operation (load follow or base load).

As indicated in Figure 17, it should be noted the similitude of the <sup>58</sup>Co total deposited activity evolutions before and after SGR : only the levels are different.
Thanks to the low <sup>60</sup>Co contamination on the SG tubes, the dose rate around the SG tube bundles is low for the first cycle after SGR (● Figure 14). It increases with the radioactive cobalt contamination and then decreases with <sup>58</sup>Co.

<figure><img src="image_15.png" /><figcaption>Figure 14: Dose rate at the SG tube bundles</figcaption></figure>

### <sup>58</sup>Co and <sup>60</sup>Co Contribution

Only <sup>58</sup>Co and <sup>60</sup>Co are taken into account to evaluate the radionuclide contribution, even if they represent about 90% of the dose rate for the reactors concerned.

Figure 15 and Figure 16 show the <sup>58</sup>Co contribution to the dose rate in the vicinity of the crossover legs and SGs, respectively.

Just before SGR, <sup>60</sup>Co contributes to 70%-80% of the dose rate at primary system. Because of SGR, the <sup>58</sup>Co contribution increases and generally becomes the major one. The <sup>58</sup>Co contribution reaches the levels preceding SGR with alloy 800 much faster than SGR with Inconel.

<figure><img src="image_16.png" /><figcaption>Figure 15: <sup>58</sup>Co contribution to the dose rate at the crossover legs</figcaption></figure>

<figure><img src="image_17.png" /><figcaption>Figure 16: <sup>58</sup>Co contribution to the dose rate at the SG</figcaption></figure>

## Discussion

### <sup>58</sup>Co contamination

Because of SGR, Nickel release increases in the primary system up to three cycles. This rise is shown in an increase of the <sup>58</sup>Co deposited activity on the primary coolant loops and an increase of <sup>58</sup>Co activity released from fuel rods during shutdown with oxygenation. Then, the <sup>58</sup>Co contamination decreases for several cycles and generally reaches lower levels than before SGR.

These levels do not seem to mainly depend on alloy composition :

* a SGR with alloy 600TT can lead to a <sup>58</sup>Co contamination similar to a SGR with alloy 690TT,

* the <sup>58</sup>Co deposited activity on the alloy 690TT SG tubes can be comparable to the one on the alloy 800 SG tubes,

* the decrease of <sup>58</sup>Co contamination, after a SGR with alloy 800, is less than expected; except on the hot side of primary coolant loops. This relative high <sup>58</sup>Co contamination shows that there is still a large source of Nickel after a SGR with alloy 800.

After a SGR, the <sup>58</sup>Co variations can be due to :

* a release and a trapping of SG tubes depending on the manufacturing process of new SG tubes [1] [5],

* a higher release of SG tubes during formation of the inner protective oxide layer,

* a « memory effect » due to Nickel deposits in the primary system before SGR, this effect decreasing with this Nickel reserve.

In any case, the contamination levels reached after SGR are not due to reactor coolant chemistry as long as pH<sub>300°C</sub> lies between 7.0 and 7.4. As there is no correlation with reactor operation (load follow or base load).

As indicated in Figure 17, it should be noted the similitude of the <sup>58</sup>Co total deposited activity evolutions before and after SGR : only the levels are different.
<figure><img src="image_18.png" /><figcaption>Figure 17 : $^{58}$Co total deposited activity on the out-of-flux surfaces of primary circuit</figcaption></figure>

$^{60}$Co contamination

Because of the $^{60}$Co radioactive half-life (5.3 years), $^{60}$Co deposited activity on the new SG tubes increases for several cycles but by a lower rate than before SGR, as indicated in Figure 18. This is due to a lower cobalt content of the new SG tubes and to the improvement of their manufacturing process. By the way, it means that SG tubes are the main source of cobalt as long as there is no excessive wear of stellite hard-facing material.

Thanks to this lower cobalt release, $^{60}$Co deposited activity on the primary coolant pipes decreases. But because of the long half-life of $^{60}$Co, the contamination levels after SGR mainly result from the state before SGR

<figure><img src="image_19.png" /><figcaption>Figure 18 : $^{60}$Co total deposited activity on the out-of-flux surfaces of primary circuit</figcaption></figure>

SGR leads to an activity distribution, on the one hand, from the SG tubes to the primary piping, and on the other hand, from $^{60}$Co to $^{58}$Co. Thus generally, the $^{58}$Co contribution becomes preponderant for some cycles,

giving way to $^{60}$Co, which accumulates because of its half-life.

Conclusion

Steam Generator Replacement influences level, distribution and contribution to the dose rate of $^{58}$Co and $^{60}$Co in PWR primary circuit. These impacts vary for the first cycles after SGR. They are dependent on contamination levels before SGR, material and manufacturing process of new SG tubes, primary system parts and radionuclides taken into account, if the other parameters are constant.

This study can be summarized as follows : corrosion product deposits in primary system before SGR and manufacturing process of new SG tubes seem to be the two main key factors for distribution and levels of contamination after SGR. Although tube material (alloys 600, 690, 800) is important, it seems to be of second order, as reactor coolant chemistry as long as pH$_{300^{\circ}$C lies between 7.0 and 7.4

The first key factor applies especially to $^{60}$Co because of its radioactive half-life. SGR with alloy 600TT can lead to a contamination comparable to the one due to SGR with alloy 690TT. And contamination after a SGR with alloy 690TT can be similar to the one after a SGR with alloy 800, alloy 800 bringing not all the expected profit.

Generally after several cycles, a SGR enables to reach contamination levels lower than before SGR.

References

[1] S. Anthoni, et al., Inventory of corrosion product behaviour on EDF's plants, Chemistry in Water Reactors Operating Experience and New Development, Nice, 1994

[2] R. Eimecke and S. Anthoni, Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC), 7th International Conference on radiation Shielding, Bournemouth, 1988

[3] S. Anthoni et al., Behaviour of the corrosion products during PWR shutdowns, Nice, 1994

[4] P. Beslu et al., Trend of plant radiation fields of French reactors : analysis and perspectives, BNES, London, 1989

[5] S. Anthoni et al., CORELE : an out-of-pile loop for corrosion product release measurement from steam generator tubes, JAIF, Japan, 1998
<figure><img src="image_18.png" /><figcaption>Figure 17 : $^{58}$Co total deposited activity on the out-of-flux surfaces of primary circuit</figcaption></figure>

$^{60}$Co contamination

Because of the $^{60}$Co radioactive half-life (5.3 years), $^{60}$Co deposited activity on the new SG tubes increases for several cycles but by a lower rate than before SGR, as indicated in Figure 18. This is due to a lower cobalt content of the new SG tubes and to the improvement of their manufacturing process. By the way, it means that SG tubes are the main source of cobalt as long as there is no excessive wear of stellite hard-facing material.

Thanks to this lower cobalt release, $^{60}$Co deposited activity on the primary coolant pipes decreases. But because of the long half-life of $^{60}$Co, the contamination levels after SGR mainly result from the state before SGR

<figure><img src="image_19.png" /><figcaption>Figure 18 : $^{60}$Co total deposited activity on the out-of-flux surfaces of primary circuit</figcaption></figure>

SGR leads to an activity distribution, on the one hand, from the SG tubes to the primary piping, and on the other hand, from $^{60}$Co to $^{58}$Co. Thus generally, the $^{58}$Co contribution becomes preponderant for some cycles,

giving way to $^{60}$Co, which accumulates because of its half-life.

Conclusion

Steam Generator Replacement influences level, distribution and contribution to the dose rate of $^{58}$Co and $^{60}$Co in PWR primary circuit. These impacts vary for the first cycles after SGR. They are dependent on contamination levels before SGR, material and manufacturing process of new SG tubes, primary system parts and radionuclides taken into account, if the other parameters are constant.

This study can be summarized as follows : corrosion product deposits in primary system before SGR and manufacturing process of new SG tubes seem to be the two main key factors for distribution and levels of contamination after SGR. Although tube material (alloys 600, 690, 800) is important, it seems to be of second order, as reactor coolant chemistry as long as pH$_{300^{\circ}$C lies between 7.0 and 7.4

The first key factor applies especially to $^{60}$Co because of its radioactive half-life. SGR with alloy 600TT can lead to a contamination comparable to the one due to SGR with alloy 690TT. And contamination after a SGR with alloy 690TT can be similar to the one after a SGR with alloy 800, alloy 800 bringing not all the expected profit.

Generally after several cycles, a SGR enables to reach contamination levels lower than before SGR.

References

[1] S. Anthoni, et al., Inventory of corrosion product behaviour on EDF's plants, Chemistry in Water Reactors Operating Experience and New Development, Nice, 1994

[2] R. Eimecke and S. Anthoni, Ensemble de Mesure et d'Etude de la Contamination des Circuits (EMECC), 7th International Conference on radiation Shielding, Bournemouth, 1988

[3] S. Anthoni et al., Behaviour of the corrosion products during PWR shutdowns, Nice, 1994

[4] P. Beslu et al., Trend of plant radiation fields of French reactors : analysis and perspectives, BNES, London, 1989

[5] S. Anthoni et al., CORELE : an out-of-pile loop for corrosion product release measurement from steam generator tubes, JAIF, Japan, 1998