A Tigeras

F Dacquait

L Viricel

J C Segura

L Guinard

J L Bretelle

A Rocher

Complete Analyse of Zinc injection impact at BUGEY 2&4

Zinc injection at BUGEY 2 and 4 provides a distinctive experience in the world, which can be used for a better understanding of the role of the zinc in the primary coolant and its impact on the radiation fields, fuel and wastes. BUGEY Unit 2 and Unit 4 are two EDF 900 MWe PWRs. The alloy 600 MA Steam Generators of BUGEY 4 have been replaced by alloy 690 TT SGs at the end of cycle 23.

EDF implemented the Zinc injection for the first time in France at BUGEY 2 Cycle 22 four years ago in order to reduce the dose rates. In 2005, the zinc injection was also initiated at BUGEY 4 Cycle 23, during the previous cycle of the steam generator replacement.

The gamma-spectrometry measurements performed at BUGEY on the steam generator tubing and primary pipes for several cycles (before and after the zinc injection) seem to show that the zinc incorporation on the oxide layers is not immediate or uniform in the different portions of the circuits and the dose rate reduction may be influenced by different factors.

The interest of BUGEY feedback is the fact that unlike other units in the world, the zinc injection is the only chemistry modification introduced in the primary coolant: other actions supposed to have a positive effect on radiation fields (pH, lithium concentration increase, fuel cleaning) have not been applied. Thereby the actual benefit due the zinc injection can be evaluated for the units similar to those of BUGEY (with alloy 600 and 690).

This paper gives an overview of the experience gathered at BUGEY during the first cycles of zinc injection and presents the future EDF strategy concerning zinc injection.

BACKGROUND

The first EDF unit injecting zinc was Bugey 2 in 2004. This unit was selected for having the highest dose rate of the French fleet. Two years later the zinc injection was also implemented in the unit 4 of Bugey with a preventive objective of minimizing core radiation fields before Steam Generator Replacement (Table 1).

The demonstration carried out at Bugey 2 and 4 comprises a complete set of analysis in order to verify the expected benefit from zinc injection as well as its innocuous effect on fuel cladding and other components of the plants.

After three cycles of injection at Bugey 2 and one cycle at Bugey 4 the results are encouraging and EDF has decided to go on with the injection in these units and extend this practice to other units taking into account the other goals of zinc addition in primary coolant: PWSCC and AOA mitigation. 1 Zinc was added for the last 6 months of cycle 22 (after checking the silica criterion, having verified the absence of AOA and preparing the injection system), the injection was interrupted for three weeks due to difficulties with the measurement device (furnace of atomic absorption spectrometry). During cycle 22, zinc was actually added for five months.

CONDITIONS OF ZINC INJECTION IMPLEMENTATION

The zinc is injected as zinc acetate depleted in 64 Zn. The zinc concentration target is 10 µg/kg with an expected range between 5 and 15 µg/kg. However, as it is described in [START_REF] Tigeras | EDF Zinc Injection: Analysis of Power Reduction Impact on the Chemistry and Radiochemistry parameters[END_REF] [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF], the load follow operation (figure 1) in EDF NPPs induces some significant variations and concentrations of about 20 µg/kg have been achieved due to the zinc return into the primary coolant.

During the first cycle of injection at Bugey 2, zinc was only added during the last 5 months due to the high silica concentration at the beginning of the cycle. The beginning of injection has been anticipated in the following cycles of Bugey 2 and 4 due to the frequent dilutions (figure 2) of primary coolant which has induced an increase on boron effluents. After analyzing the feedback of three cycles of injection, EDF and AREVA have performed a fuel evaluation in order to optimize the practice and minimize the possible operating restrictions linked to boron effluents. Bugey 2 and Bugey 4 are low duty plants with low concentrations in Ca, Mg and Al and without experience in AOA. Under these conditions, the international feedback and the experimental/theoretical results indicate that the AOA risk is negligible if the nickel concentration is not significant and the stability conditions are reached before beginning of zinc injection. Taking into account these elements, the specifications applied to Bugey 2 and Bugey 4 have evolved in conformity with the recommendations of AREVA for Low Duty Plants:

-limit nickel concentration 6 µg/kg with a target value of 3 µg/kg -limit silica concentration of 1.5 mg/kg . Once the zinc injection has started during the different cycles no difficulty has been reported. In total 8.5 kg Zn was added during the three first cycles of injection at Bugey 2 with a total exposure of 284 ppbmonths (Table 2). The zinc incorporated into the Reactor Cooling System (RCS) surfaces is commonly estimated [START_REF]Pressurized Water Reactor Primary Water Zinc Application Guidelines[END_REF] as the difference between the injected mass and the removed mass. Even if this simplified calculation can be applied for the units without load following, for EDF units this result should be considered only as an indicative maximum (Table 3). The zinc losses during power reductions seem to have an influence on the zinc mechanism incorporation [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF].

REACTOR COOLANT DATA

Cycle monitoring

The 58 Co and 60 Co activity is plotted with the zinc concentration in the primary coolant of Bugey 2 (Figure 3a). Zinc is detected when [Zn]>1 µg/kg, during the first cycle of injection after 35 days of injection and reaches the target concentration (10 µg/kg) after injecting 1800 g of zinc. The 58 Co and 60 Co activities evolutions are well correlated with the progress of zinc concentration in the primary coolant. The 60 Co becomes higher than the detection limit and reaches values of 8 MBq/t at the end of cycle 22, while the 58 Co evolution is quite similar to zinc evolution and their respective variations are a consequence of power variations. During cycles 23 and 24, the zinc reached the target concentration earlier than during cycle 22: after having injected 588 g and 429 g of zinc during cycles 23 and 24 respectively, the zinc concentration was approximately 10 µg/kg and the initial cobalt level was already higher than that observed in the previous cycle before injecting zinc.

At the end of cycles 22 and 23, the radiocobalt activity increases faster than during the cycle without any modification of zinc rate injection and zinc concentration of the zinc acetate solution (Figure 3a). In fact, this behaviour is not linked to the zinc injection but to the power reduction during stretch-out of cycle. During this period the usual activity increase is due to the release of corrosion products from fuel crud [12][13]. At this point it is relevant to indicate that the activity reached in 58 Co at Bugey 2 and Bugey 4 stays lower than the typical values measured in other units among EDF NPPs. At the end of cycle 24, it is not observed an acceleration in the evolution of radiocobalt because of the absence of stretch-out period. Similar evolutions of reactor coolant of reactor radiochemistry were observed at Bugey 4 during the 1 st cycle of zinc injection (Figure 3b).

Nickel concentration remains between 2 and 4 µg/kg during stable operation. At the end of the cycles, during stretch-out, nickel concentration slightly increases to concentrations about 5 µg/kg. During this phase of demonstration, iron has was monitored. Its concentration stayed lower than the detection limit (1 µg/kg) during steady operation and became measurable during stretch out (about 5 µg/kg).

Shutdown Monitoring

The first cycle of injection at Bugey 2 was characterized by a significant increase in the 58 Co peak at oxygenation (Figure 4a). Even if during the first cycles of injection an increase in the activities was expected according to the international feedback [3] [4], this important peak seemed to be also due to other factors: the increase of the cycle length (from 12 months to 16 months), purification difficulties during some phases of the shutdown and fast [H 2 ] decrease. After the 3 rd cycle of injection the activity values became comparable with the values measured before the beginning of the injection. Concerning the 60 Co, the peak activity was also significantly higher and stayed after cycle 24 (Figure 4b, 4c), three times higher than the values measured in cycles prior the beginning of zinc injection.

The feedback from Bugey 4 (Figure 4d) confirms the relevant evolution in 60 Co activity (3050 MBq/t). As expected, the increase in 58 Co peak (168000 MBq/t) was also observed at Bugey 4, but it was not as high as at Bugey 2. This seems to confirm the influence of the other factors identified at Bugey 2 cycle 22 on 58 Co activity Table 4).

The maximum obtained values of nickel and iron at oxygenation are not abnormal compared with other units of EDF fleet without zinc. They remain globally similar after three cycles of injection: about 4500 µg/kg for nickel and 92 µg/kg for iron. Nevertheless, according to the international feedback a reduction of nickel release during the shutdown is expected after some cycles of injection [START_REF]Pressurized Water Reactor Primary Water Zinc Application Guidelines[END_REF].

The most important difference between Bugey feedback and international experience is the release of zinc observed during the oxygenation process. Figure 4 represents the monitoring of zinc during the shutdowns. The peaks obtained at Bugey 2 (cycles 22 and 23) and Bugey 4 (cycle 23) are twice as high as the highest peak reported until now (300 µg/kg). This significant release may be explained by differences in the core design with the other units or/and by the influence of load following during the cycles. It should be noted that during the shutdown of Bugey 2 cycle 24 the peak had significantly decreased to 60 µg/kg.

FUEL SURVEILLANCE

During the cycle the flux maps did not evidence any indication of AOA and no operating restrictions were identified.

The cores of Bugey 2 and 4 are loaded with Zircaloy 4 fuel assemblies (F/AS) from AREVA.

The fuel performance during the first three cycles of injection at Bugey 2 was evaluated through visual inspections and thickness oxide measurements for 3 rd cycle assemblies with the highest burn-up in the core.

Visual examination did not revealed any particularity of these F/AS when compared to F/AS irradiated without zinc injection. Moreover, the oxide thickness measurements performed during the three cycles of demonstration are completely comparable with the data compiled in the past without zinc. Fuel oxide measurements also agreed well with predicted values for Zircaloy cladding.

Based on these results, the absence of acceleration of corrosion on fuel cladding is confirmed and the current model seems to be also qualified to predict the oxide evolution for units with zinc (Figure 5). MWe units without zinc injection.

EFFLUENTS AND WASTE SURVEILLANCE

The impact of zinc acetate on the environmental has been analyzed by EDF during the demonstrations of Bugey 2 and Bugey 4.

The temporary cobalt activity increase in primary coolant owing to the zinc action onto the oxide surface may have a potential impact on plant waste generation: liquid waste effluents, spent filters and resins. The data from Bugey 2 and 4 confirm the international feedback synthesized in the reference [START_REF]Pressurized Water Reactor Primary Water Zinc Application Guidelines[END_REF]. The radiocobalt increases in primary coolant during the cycle and shutdown do not necessarily imply an increase in the plant radioactive liquid effluents if the correct purification systems operation is assured (figure 6 and7). Moreover, the use of zinc depleted in 64 Zn enables to ensure the absence of 65 Zn in the liquid wastes.

As a consequence of the correct retention of Cobalt by the demineralizers, after the first cycle of injection at Bugey 2, an increase of the percentage of 60 Co was observed on the filters. Nevertheless, during the following cycles, the radionuclide percentages have been comparable to the standard spectrum. The activities of solid wastes generated during the zinc cycles did not exceed the limit values for guarantying their evacuation.

EDF has also performed an exhaustive analysis of long life radionucleides (RN) in the primary coolant ( 55 Fe, 63 Ni, 94 Nb, 14 C, 90 Sr) in order to ensure that the typical ratios (typical ratio= RN/ 60 Co) used by EDF to characterized the solid wastes are still relevant in the presence of zinc. This has been confirmed by the results from Bugey 2 and Bugey 4.

DOSE RATES EVOLUTION AND SURFACE CONTAMINATION

Experimental results

The doses rates (Figure 8) measured in the different points of circuits have been correlated with the surface activities (Figure 9) given by gamma spectrometry (EMECC device) [START_REF] Eimeke | Ensemble de Mesures et d'Etude de contamination des Circuits[END_REF] performed by the CEA in order to quantify the actual effect of zinc injection on contamination. These contamination measurements have been carried out in the French fleet since 1971. The experience acquired during this period of time in more than 270 campaigns permit to give a high level of confidence to the results.

The contamination at Bugey 2 was frequently monitored in the past, which helps to better identify the possible particularities of Bugey 2 after Zinc injection.

Between cycles 21 and 24 it should be noted: • The 60 Co surface activities are practically constant between cycles 21 and 23. Between cycles 23 and 24 a diminution of 12-14% is reported on all the portions of circuits. This decrease corresponds to the natural decay of 60 Co and consequently it might be assumed that zinc has only contributed to avoid its incorporation on the ex-core surfaces. These diminutions are sometimes observed in other units without zinc; they are, however, only detected in some points of measurement and/or they are only about 5-10%.

• There is not a general tendency for 58 Co surface activities on the primary system between cycles 21 and 24. On the other hand, between cycles 23 and 24, they are constant on the primary system. The evolution of cobalt activities between cycle 23 and 24 is relatively well correlated with the evolution of doses rates (Figures 89) considering their higher uncertainties (at least 30%):

• On cross-over piping and cold legs the decrease is more significant after cycle 24.

• On hot legs there was a tendency to increase from cycle 20. This trend was reversed after cycle 23 with a diminution of 15% following with a more significant decrease of 30% after cycle 24.

At Bugey 4 only one cycle of injection has been performed. A 60 Co activity average diminution of 14% has been detected on cross-over piping and steam generators with a 18% decrease of dose rates. In the other parts of the RCS, no significant evolutions have been detected either in deposited activity or in dose rates.

Discussion and Interpretation of results

After 27 months of injection, the positive effect of Zn on surface activities seems to be detected at Bugey 2. Based on the reduction of 60 Co surface activity (similar to the natural decay), it is supposed that only the 1 st mechanism of inhibition [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF] probably occurred at Bugey 2. The presence of Zn has probably had a positive influence to avoid a new recontamination onto the surface of the circuits. At Bugey, up to now the Zn may only reduce the pickup of 60 Co, since the feedback from Bugey does not permit to suppose that the Zn diffusion process into the inner oxide layers (observed in other units [START_REF]Pressurized Water Reactor Primary Water Zinc Application Guidelines[END_REF]) has begun after 3 cycles of injection. Therefore, it is not possible to affirm that the mechanism of Co substitution has occurred in the deepest layers of oxide in Bugey.

Comparing the surface activity measurements from Bugey with the measurements carried out by CEA in German units [START_REF] Sttellwag | Zinc Injection in German Plants[END_REF][8], some aspects can be underlined during the first cycles of injection for these units. Figure 9 represents the average activity evolution of the primary system contamination (SG tubes, hot, cold or crossover legs) during the first cycles of injection at Bugey 2, 4 and Biblis B. From German measurements, the following aspects can be underlined:

-the feedback from Biblis B indicates a slow and progressive reduction in 60 Co activity. The decrease rate almost corresponds to the radioactive decay. However, we can wonder if the decrease would not have been the same without zinc injection. Indeed, the 60 Co contamination level in the primary pipes of Biblis-B was similar, or even higher, than the level at BUGEY-2 and, as observed in Figure 9, the 60 Co contamination had decreased at BUGEY-2 without Zn injection (i.e. before cycle 22). As already explained in 2004 [START_REF] Dacquait | Status of primary system contamination in French PWRs[END_REF], once the abnormal cobalt source was removed (replacement of Inconel 718 fuel spacer grids with increased cobalt content in BUGEY-2 and BIBLIS-B [START_REF] Haag | Operational Experience with Zinc Addition in a German PWR[END_REF]) or exhausted, the 60 Co contamination decreased significantly by a factor of between 2 and 4 over roughly 10 years2 in BUGEY-2 (between cycles 12 and 20) on the primary coolant pipes and this, without zinc injection. It seems that the 60 Co build-up coming from the remaining sources is low or even negligible compared to the former 60 Co contamination on the primary coolant pipes.

-EMECC measurements were also performed at Obrigheim. They did not show a 60 Co contamination reduction after 2 cycles of Zn injection.

This comparison of surface activity measurements and dose rates evolution during the first cycles of zinc injection in German plants seems to indicate a similar behaviour with Bugey feedback, although the aimed Zn content in German PWR primary systems is 5 ppb whereas it is 10-15 ppb in French PWR primary systems. The effect of zinc seems to be slower than in other units in the world. According to the analysis carried out in [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF] this behaviour can be due to:

the age of the units: in fact, Obrigheim, Biblis B, Bugey 2 and 4 are up to now the units where the Zn injection was implanted later in the life of the plant. The thickness of the oxides on RCS surfaces can play a crucial role on the zinc efficiency.

-Power variation perturbations: the zinc injection at German units with frequent mi-cycle shutdowns and power reductions show also similarities with activity and dose rates evolutions from Bugey [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF].

In spite of not having obtained the same effects than the majority of the units in the world, the 60 Co activity evolution measured on the surfaces seems to be correlated correctly with the dose rates evolution and the first benefits started to be observed after cycle 24 of Bugey 2. According to the global experience of German plants [START_REF] Tigeras | Understanding the zinc behavior in PWR primary coolant: a comparison between French and German experience[END_REF] [8] and of the other units in the world, a progressive reduction can be expected for the future cycles of Bugey. To make sure of the favourable impact of Zn on the 60 Co contamination, it is necessary to confirm the decrease in 60 Co of about 12% after the next cycles of Bugey-2 and after the 1 st shutdown of Bugey 4 which has the particular interest of being the first cycle following SGR. This shutdown is planned in July 2008, thereby it is not yet possible to evaluate the zinc injection advantages at Bugey 4. Bugey 2 and Bugey 4 have been selected for the zinc demonstration due to their high levels of dose rates in primary coolant. However, the zinc injection presents also an interest for units where decontamination operations of auxiliary circuits are programmed. In fact the international experience, from units injecting zinc after decontamination operations, shows a diminution of the recontamination rate. This can be linked to the zinc contribution to reduce the re-adhesion of 60 Co onto the surface of primary coolant and to diminish the potential source term which can be transferred to auxiliary systems. The zinc injection may be also then considered as a preventive method of reducing the contamination of auxiliary systems in units with high levels of doses due to 60 Co [START_REF] Dacquait | Auxiliary System Contamination in French PWRs[END_REF].

CONCLUSION AND EDF STRATEGY

The demonstrations of Bugey 2 and 4 allowed to verify some aspects from international feedback and to identify some particularities of EDF NPPs operation.

The first objective of the demonstration has been reached: the innocuous effect of zinc on components, materials, effluents, wastes and fuel was verified.

In terms of benefits, the results of the demonstrations at Bugey 2 and Bugey 4 shows similar dose rates and surface activities evolutions than the German experience during the first cycles of injection. After the third cycle of injection at Bugey 2 the results are encouraging but need to be confirmed. Therefore, it is expected to obtain at Bugey similar dose reductions than that observed in German plants following the first cycles of injection and in other plants in the world.

Based on the Bugey results and on the international feedback EDF has decided in 2008 to implement the zinc addition on other units than Bugey 2 and Bugey 4, not only to control contamination and dose rates but also to manage a potential AOA risk on future core designs [START_REF] Decossin | Axial Offset Anomaly Feedback and R&D modelling studies on EDF cores[END_REF] and to mitigate PWSCC. The EDF strategy is illustrated in the diagram of Figure 10. According to this strategy, zinc injection is planned to begin in about 9 more EDF units from 2009 to 2011.

Technical Specifications are been reviewed by EDF with the fuel vendors especially for the progressive implementation of high zinc concentrations up to 39-40 μg/kg for PWSCC mitigation purposes.

At present time, EDF is developing the Guidelines, recommendations and procedures to facilitate in its units the application of zinc injection and the monitoring of parameters request in order to improve the interpretation of results. .

Figure 1 :

 1 Figure 1: Zinc Evolution in RCS during load follow periods.

Figure 2 :

 2 Figure 2: SiO 2 evolution in RCS during the cycles of zinc injection

Figure 3a :

 3a Figure 3a: Correlation between [Zn] and radiocobalt activity evolution in RCS ( 58 Co and 60 Co) at Bugey 2.

Figure 3b :

 3b Figure 3b: Correlation between [Zn] and radiocobalt activity evolution in RCS ( 58 Co and 60 Co) at Bugey 4.

Figure 4a :

 4a Figure 4a: Radiocobalt and Zinc evolution during shutdown of Bugey 2 cycle 22.

Figure 4b :

 4b Figure 4b: Radiocobalt and Zinc evolution during shutdown of Bugey 2 cycle 23

Figure 5 :

 5 Figure 5: Measurement of Fuel Oxide Thickness at Bugey 2 with zinc injection compared with data from 900MWe units without zinc injection.

Figure 6 :

 6 Figure 6: 58 Co contribution in liquid effluents at Bugey

Figure 7 :

 7 Figure 7: 60 Co contribution in liquid effluents at Bugey

Figure 8a. 2 :

 2 Figure 8a.2 : Dose rate around SG shell

Figure 10 :

 10 Figure 10 : Zinc Injection Strategy at EDF

Table 1 :

 1 History of Zinc Injection at Bugey 2 and 4

	Unit	Cycle	Beginning of	End of cycle	Start of Zn	Zn	Injection
			cycle		Injection	Duration
	BUG2	22	28/10/03	18/02/05	07/09/04	5 months 1
		23	27/03/05	10/06/06	22/06/05	12 months
		24	01/08/06	11/08/07	26/09/06	10 months
		25	10/09/07	Cycle in progress	14/01/08	Cycle in progress
	BUG4	23	01/10/05	24/12/06	24/01/06	11 months
		24	26/06/07	Cycle in progress	25/07/07	Cycle in progress

Table 2 :

 2 Zinc Injection Evolution

	Unit	Cycle [Zn]average	Exposure
			(µg/kg)	(ppb_months)
	BUG2	22	5	25
	BUG 2	23	9.7	116
	BUG 2	24	13	143
	BUG 4	23	5.9	67.8

Table 3 :

 3 Estimation of Zinc Incorporated onto RCS Surfaces

	Unit Cycle Injection	Zinc	Zinc	Estimated
			Duration	injected	removed	zinc mass
			(months)	Mass	mass (g)	incorporated
				(g)		onto RCS
						surfaces (g)
	BUG2	22	5	2515	264	2251
	BUG2	23	12	4020	1360	2660
	BUG2	24	10	1970	479	1491
	BUG4	23	11	4870	550	4320

Cycle 23, from BOC: 23/March/2005 to April 2006 (EOC: 10/June/2006)

		25																																																100
																																																		90
		20						Start																																									80
								Injection
																																																		70
	Zn Concentration (ppb)	10 15																																																40 60 50	%Pn
																																																		30
		5																																																20
																																																		10
		0																																																0
			7	17	25	36	47	56	68	78	88	96	103	110	115	121	126	129	133	139	144	150	158	165	172	183	192	199	207	215	221	230	239	248	255	262	272	282	290	297	306	315	325	332	338	345	352	360	369	374	380
																									Days from BOC
																					Zinc Injection at Bugey 222
																									Pow er %			Silica (µg/kg)
																																																		3000
		100
																																																		2750
			90																																															2500
			80																																															2250
	%Power	40 50 60 70																																															1000 1250 1500 1750 2000	Silica (µg/kg)
			30
																																																		750
			20																																															500
			10																																															250
			0																																															0
			16-sept-03		6-oct-03	26-oct-03	15-nov-03	5-déc-03		25-déc-03		14-janv-04		3-févr-04	23-févr-04	14-mars-04		3-avr-04		23-av r-04		13-mai-04		2-juin-04	22-juin-04	12-juil-04		1-août-04		21-août-04		10-sept-04	30-sept-04	20-oct-04	9-nov-04		29-nov-04		19-déc-04		8-janv-05	28-janv-05	17-févr-05	9-mars-05	29-mars-05
																											Date

Silica Specification:1ppm Beginning of Injection

Table 4 :

 4 Cobalt activity evolution at Oxygenation during shutdowns before and after Zinc injection at Bugey 2

BUG224_3 vs 900 MWe NPPs feedback

										100%	Co58% in Liquid Effluents
										90%	Bugey
										80%	Average 900 Mwe Min of 900 MWe Units	Start Zn Injection BUG4
										70%	Max of 900 MWe Units
										60%
										50%	Start Zn Injection BUG2
										40%
										30%
										20%
										10%
										0%
										2002	2003	2004	2005	2006	2007
		140
		120
	Maximum average thickness (µm)	40 60 80 100							900 MWe feedback FX20L4 FX20H4	50% 60% 100% 70% 80% 90%	Co60 % in Liquid Effluents Start Zn Injection Bugey Average 900 Mwe Min of 900 MWe Units Max of 900 MWe Units	Start Zn Injection BUG4
										40%	BUG2
		20								30%
		0								20%
		0	10000	20000	30000	40000	50000	60000	70000	10%
					Fuel rods burnup (MWd/t)				0%
										2002	2003	2004	2005	2006	2007

60 Co decay leads to a contamination decrease of about 4 over 10 years.

ACKNOWLEDGEMENTS

The zinc injection analysis at Bugey and the preparation of its application in other units are the result of a cooperative work carried out by different groups of EDF. The collaboration of following persons should be highlighted:


Luc Guinard
email: luc.guinard@edf.fr

Chris Bates

Gavin Lancaster

Frédéric Dacquait

Characterization of contamination : EMECC campaign at Sizewell B and comparison with French PWRs

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

INTRODUCTION

The reduction of collective dose is an important objective for all nuclear plant operators. One of the support for studying the impact of design and operation parameters is the measurement campaigns of contamination in PWRs. EDF performs consequently gamma spectrometry measurements with CEA and his EMECC system [1], with more than 200 campaigns in EDF plants and also about 70 campaigns in collaboration with other operators. The comparison on low level contamination plants give elements to improve the behaviour of the plants. Consequently British Energy and EDF collaborated recently for measurements during Refueling Outage at Sizewell B.

EMECC CAMPAIGN AT SIZEWELL B

Plant measurements

Sizewell B is a Westinghouse 4-loops PWR with alloy 690TT tubes steam generators (SG). This unique British PWR is characterized by very low corrosion product releases and radiation fields [2]. In order to compare the contamination level of Sizewell B to that of French PWRs, an EMECC campaign was performed during the refueling shutdown at the end of cycle 8.

The reactor coolant chemistry is boron-lithium coordinated chemistry with a target pH 300°C of 7.2. As a difference with French use, lithium can be over 2.2 mg/kg, till 3.5 mg/kg. The H 2 target concentration ranges between 30 and 45 cm 3 /kg, slightly higher than French range (25 -35 cm 3 /kg). Cycle 8 lasted 475 EFPD, with the various steps of shutdown presented in figure 1. All the results will be compared with measurements performed at Saint-Laurent B1 during the refueling shutdown at the end of cycle 20. Indeed, Saint Laurent B1 is a Framatome 3-loops PWR with alloy 690TT tubes SG and equivalent duration of operation since SG replacement (performed after cycle 11).

Deposited activities

The deposited activities are presented in figure 2 for 58 Co in RCS, figure 3 for 60 Co in RCS, figure 4 for CVCS letdown pipe and figure 5 ). The contribution of 58 Co and 60 Co is higher than 90% : 70% to 85% for 60 Co, 10% to 20% for 58 Co.

For CVCS, dose rate measurements are presented in table 2. For letdown pipe, the 58 Co contribution is higher : 30% compared with 60% for 60 Co due to the thinner pipe thickness. 60 Co is the major contributor to the dose rates around the letdown heat exchanger (75%), with 110m Ag and 58 Co (about 7% each). The comparison with St-Laurent B1 for RCS shows similar dose rates whereas 60 Co surface activities are higher in SIZB08. Sizewell has the benefit of thicker component walls (e.g. SG shell thickness: 9.7 cm in SIZB compared with 8.5 cm in SLB1; hot leg thickness: 8.3 cm in SIZB compared with 6.9 cm in SLB1).

For CVCS letdown pipe, the dose rate are higher at SIZB08, due to higher surface activities and thinner walls. Inversely for letdown heat exchanger, the dose rate are lower in SIZB08, due to high 110m Ag contamination in SLB120.

Volume activities

COLLECTIVE DOSIMETRY

The comparison between Sizewell B and St-Laurent B1 (see figure 8) indicates that Sizewell B has slightly lower collective dosimetry. However, St Laurent B1 performed a significant improvement for the last decade, due to efficient ALARA plans.

The average values for N4 units are very similar with Sizewell ones. N4 units have also significantly lower dose rate index than St-Laurent B1, which has the same trends than Sizewell B.

It can therefore be concluded that if the dose rates are very comparable (Sizewell B and St-Laurent B1), the collective dosimetry is slightly lower for Sizewell B. French units have similar collective dosimetry to Sizewell B one only with lower dose rates.

DISCUSSION

The evidenced differences between Sizewell and Saint-Laurent are now to be discussed, to find explanations and ways of improvement.

Design

The potential reasons for higher 60 Co contamination in SIZB08 are reviewed :

-It should not be the SG tubing Co content , because it is very low (< 150 ppm) either in SIZB and SLB1.

-The influence of stellites is also not probable, because there is stellites reduction in CVCS valves in SIZB, not in SLB1.

-Inconel 718 grids could be the cause : for the first cycle in SIZB, all assemblies were equipped with Inconel 718 grids. A major part of them has been replaced, but there were still 13 assemblies equipped with Inconel 718 grids during cycle 8 (and 0 for SLB1). These grids have a specified maximum Co level of 400 ppm. Even if the majority have been now removed, their initial use can contribute to higher 60 Co contamination for many cycles because of 60 Co radioactive half-life (5.3 y).

Nevertheless, this cause is not proved. Inconel 718 grids (52% of Ni) should also increase 58 Co contamination, and this phenomenon is not obvious.

Both units have in common low 58 Co surface and volume activities several cycles after commissioning or SGR. One cause of this low 58 Co contamination level can be the very low corrosion release rate of alloy 690 SG tubing, due to improved manufacturing process.

Operation

There are no obvious difference in operating parameters between Sizewell and Saint-Laurent. The main differences are the longer cycle and the higher lithium content at beginning of cycle for Sizewell, but these parameters are not known for changing significantly the contamination [3].

Another cause for the low 58 Co surface and volume activities several cycles after commissioning or SGR should be Hot Functional Test (HFT) passivation procedure. However, there has been no passivation procedure in SLB1. As 58 Co peaks is below 10 GBq/t 6 cycles after SIZB commissioning and 8 cycles after SLB1 SGR, it indicates probably that a modified passivation procedure with a complete water purification stage like in SIZEWELL-B should lower contamination level more quickly.

Last, we can note that volume activities are quite different, with lower values for SIZB, whereas similar volume activities during shutdowns and similar surface activities. For explaining such differences, maybe without impact on contamination, some exchanges between EDF and BE are in progress.

Other causes

An interesting indicator could be the time spent in controlled area. It is surprisingly higher for Sizewell B outages (about 80 000 h) than for French units outages (about 40 000 h). At first analysis, it could seem contradictory with "logical" and international trends [3] : the reduction of time is one of the main contributors for RP improvement. This data amplify the difference between Sizewell B and St-Laurent B1 : with similar dose rates and more time spent in controlled area, the collective dosimetry is lower for Sizewell B.

This analysis has consequently to be completed, by examining the differences in organization. It appears firstly that people work less "alone" during Sizewell B outages, with almost systematically one Health Physician together with each operator. This organization has two direct consequences : it can improve the working conditions and it increases "artificially" the time spent in controlled area. Another possible difference is less time in dosing outage operations, as for instance less work for SG inspections.

Other potential causes of differences have to be evaluated :

differences in contamination on auxiliary systems, even if first results on CVCS (See figure 4 and 5) are not conclusive. -differences in volume activities : could higher activities in French reactors have an indirect impact on contamination ? (it is well known that there is no direct impact, as more than 90% of dose rates are due to surface activities).

PERSPECTIVES

To explore further these hypothesis, complementary exchanges are in progress between BE and EDF. It should include measurements on auxiliary systems, accurate comparison for chosen outage operations and further investigations on differences in volume activities.

All these comparisons will give new elements to BE and EDF to improve the behaviour of their own plants.

Figure 1 .

 1 Figure 1. Cold shutdown procedure -Sizewell B RF08

Figure 3 .Figure 4 .

 34 Figure 3. 60 Co RCS surface activities -Comparison Sizewell B / Saint-Laurent B1.

Figure 5 .Figure 6 .

 56 Figure 5. Letdown heat exchanger surface activities -Comparison Sizewell B / Saint-Laurent B1.2.3. Dose rates

Figure 7

 7 Figure 7 compares the 58 Co and 60 Co volume activities (including soluble and insoluble parts) for Sizewell B08 and Saint-Laurent B120. Excepting the two first operating months, the volume activities are 6 to 8 times lower in Sizewell B. SIZEWELL B08 / ST LAURENT B 120 -Volume activities during cycle

Figure 8 .

 8 Figure 8. Comparison of collective dosimetry for Sizewell B, St-Laurent B1 and N4 units.

ycles (from S GR for S LB 1) GB q/m² SLB 1 SIZB Crossov e r le g -58 Co de posite d activ ity

  for letdown heat exchanger.

		Hot leg -	58 C o dep osited activity
	6
	4
	2
	0
	0	2	4	6	8	10
	5 10 C 0 15
	0	2	4	6	8	10
		C

ycles (from S GR for SLB 1) GB q/m² S LB1 S IZB SG tubing - 58 Co deposited activity

		1.5
	GBq/m²	1				Cold side	SIZB Série2 SLB1
		0.5
						Hot side
		0
		0	2	4	6	8	10
				Cycles (from SGR for SLB1)
	Figure 2. 58 Co RCS surface activities -Comparison
			Sizewell B / Saint-Laurent B1.
	58 Co and 60 Co contribute for more than 90% of the RCS
	and CVCS letdown pipe dose rates. The hot side of SG
	tubing is less contaminated than the cold side, either for
	58 Co or	60 Co. The only main other nuclide for dose rate
	contribution is 110m Ag for CVCS letdown heat exchanger
	(about 7%).

  The comparison with St-Laurent B1 shows that the 58 Co surface activities are practically identical, as the 60 Co surface activities on hot legs. The

					60 Co surface
	activities on crossover legs and SG tubing is higher in
	Sizewell B.
	The letdown pipe is more contaminated in SIZB08
	by the 2 main contributors (	58 Co and	60 Co). Inversely, the
	letdown heat exchanger is more contaminated in SLB120
	by the 3 main contributors (	58 Co,	60 Co and	110m Ag).

Table 1 .

 1 Dose rates around primary system (CEA index -mSv/h) -Comparison Sizewell B / Saint-Laurent B1.

		SIZB08	SLB120
	Primary piping	0.19	0.18
	SG tube bundle	0.10	0.08

Table 2 .

 2 Dose rates at CVCS (mSv/h) -ComparisonSizewell B / Saint Laurent B1.

		SIZB08	SLB120
	Letdown pipe	2.0	0.8
	Letdown heat exchanger	0.15	2.3


Y Nakamura
email: yuki.nakamura@latmos.ipsl.fr

LATMOS F Leblanc

N Terada

I Murata

H Nakagawa

S Sakai

S Hiruba

R Kataoka

K Murase

Numerical prediction of changes in atmospheric chemical composition by precipitation of solar energetic particles at Mars

Introduction:

Solar energetic particles (SEPs) are highenergetic particles that consist mainly of electrons and protons with energies from a few tens of keV to GeV ejected from the Sun associated with solar flares and coronal mass ejections. SEPs that precipitate into planetary atmospheres cause ionization, excitation and dissociation of atmospheric molecules, leading to changes in atmospheric chemical composition via chemical network (e.g. Solomon et al., 1981;Adams et al., 2021). The effects of SEPs on the atmospheric chemical composition on Earth has been intensively studied. For instance, during the enormous solar flare that occurred in late October 2003, NOx (NO + NO 2 ) and HOx (OH + HO 2 ) concentrations were enhanced and ozone concentration was depleted by 40% at the polar lower mesosphere (e.g. Jackman et al., 2005). Recently, the Mars Atmosphere and Volatile EvolutioN (MAVEN) spacecraft has discovered global diffuse aurora on the nightside of Mars during SEP events (Schneider et al., 2015), which has been shown to be generated mainly by SEP protons (Nakamura et al., 2022). However, changes in the neutral chemical composition of the Martian atmosphere during SEP events have not yet been estimated. By coupling a Monte Carlo model, Particle TRansport In Planetary atmospheres (PTRIP) (Nakamura et al., 2022) and a newly developed 1-dimensional photochemical model, we investigated the effects of SEPs on the atmospheric composition at Mars.

Models:

In this study, we used PTRIP to calculate production rate of CO 2 + and N 2 + in the Martian atmosphere under a SEP proton flux taken from a SEP event that occurred on 28 October 2003 at Earth (Mewaldt et al., 2005). Production rates of N and N( 2 D), which are important drivers for NOx chemistry, were estimated assuming a ratio N 2 + :N:N( 2 D) = 1:0.84:1 (Krasnopolsky, 2009). Ionization rate by cosmic rays as an ionization source at low altitudes were taken from Haider et al. (2009). In the 1-dimensional photochemical model, we implemented 486 chemical reactions consisting of HOx chemistry (Chaffin et al., 2017), NOx chemistry (Nair et al., 1994), ionospheric chemistry (Fox and Sung, 2001) and water cluster ion chemistry (Molina-Cuberos et al., 2001;Verronen et al., 2016) to fully describe chemical network during SEP events.

Results:

We first ran the photochemical model only for neutral species to obtain the steady state neutral composition. After that we ran the model for 10 days including ion species with solar extreme ultraviolet flux and cosmic ray as ionization sources to obtain the quasi-steady state solution for ion profiles. Then we ran the model with additional ionization source of SEP protons for 1 day to examine the changes in ion and neutral compositions during SEP events.

The calculated ion profiles before SEP event and after 1 day of SEP event onset is shown in Figure 1. Ion and electron densities were enhanced below 100 km altitude and densities of water cations were enhanced below 70 km altitude. The results indicate that ionization by SEP protons is more effective than ionization by cosmic rays, producing large amount of water cations below 70 km altitude, which agree well with the previous modeling study by Sheel et al. (2012). The change in neutral chemical composition was investigated. The calculated density profiles of ozone, HOx and NOx before SEP event, 1day after SEP event onset, 1 day after SEP event termination and 10days after SEP event termination are shown in Figure 2. Enhancement of HOx and decrease of ozone by a factor of 10 occurred in the altitude range of 20-60 km and enhancement of NOx by a factor of more than 100 occurred in the altitude range of 20-100 km 1 day after SEP event onset. After the end of the SEP event, ozone and HOx densities were almost recovered 1 day after the end of SEP event, however, NOx density was not recovered even 10 days after the end of SEP event, due to the different loss time scales between HOx and NOx. These results indicate that, as on Earth, similar effects on ozone, HOx and NOx concentrations are expected even in the case of the CO 2 -dominated Mars atmosphere.

Figure 2. Calculated density profiles of ozone, HOx (OH + HO 2 ) and NOx (NO + NO 2 ) before SEP event, 1 day after SEP event onset (represented as "during SEP"), 1 day after SEP event termination (represented as "recovery 1 day") and 10 days after SEP event termination (represented as "recovery 10 day").

Figure 1 .

 1 Figure 1. (upper) Calculated ion profile before SEP event and (bottom) after 1 day of SEP event onset.


F Dacquait

F Nguyen

H Marteau

B Larat

D You

G Plancque

L Guinard

G Ranchoux

C Viala

M Bachet

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

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

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

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

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

INTRODUCTION*

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

The PACTOLE code has been developed by the CEA in collaboration with EDF and AREVA NP for more than 30 years [START_REF] Beslu | A computer code PACTOLE to predict activation and transport of corrosion product in PWR[END_REF], [START_REF] Beslu | PACTOLE-PROFIP : two codes allowing prediction of the contamination of PWR primary circuits[END_REF], [START_REF] Tarabelli | Status and future plans of the PACTOLE code predicting the activation and transport of corrosion products in PWRs[END_REF], [START_REF] Marchetto | PACTOLE V3 : a new code version to predict corrosion product contamination[END_REF], [START_REF] Nguyen | First numerical simulations of contamination of the PWR primary circuit by † PROFIP : simulation code of fission product and actinide transfer in the PWR primary system[END_REF]. It is designed to simulate the transfer of corrosion products in the primary system of a PWR. It determines the space-time distribution of masses and/or activity levels of corrosion products and activation products in both solid and liquid phases in the primary system.

This paper aims to describe the modelling used to simulate the transport of corrosion products as ♣ Present affiliation : Euriware implemented in the PACTOLE V3.2 code. It also provides comparisons between simulations and experimental measurements performed in PWRs.

PWR CONTAMINATION PRINCIPLE

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

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

The primary coolant transports the ions produced by corrosion and release phenomena or by the dissolution of oxides. When the coolant becomes oversaturated in corrosion products (depending on the local temperature and chemical conditions), the ions either precipitate on the surface of the walls, or precipitate to form particles in the coolant. Particles are also produced by the erosion of deposits. When transported by the primary coolant, these particles can agglomerate and deposit on the primary system walls.

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

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

MODELLING OF THE PACTOLE V3.2 CODE

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

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

The primary system is composed of a main line (blue volumes in Figure 1) to which various lines and circuits are connected, e.g. the chemical and volume control system (CVCS). The loops of a PWR primary system are generally considered as one main loop (see Figure 1) but it is possible to differentiate between them to take into account, for example, the state of a different surface of the steam generator tubes. Seven media can be taken into account in each control volume (see Figure 2): the base metal, the inner oxide layer, the outer oxide layer, the deposit, the ions and particles carried by the coolant, and the filters for particles and ions. Base metals: alloys composing the different elements of the primary system. The metal in contact with the primary coolant is subjected to the corrosion process.

Inner oxide layer: described as a rather adhesive layer composed of small-sized grains. This oxide coats all surfaces in contact with the primary coolant. Its chemical composition is rich in chromium. When coating stainless steels and nickel-based alloys, this oxide is a chromite according to the following formula: Co x Ni y Fe 1-x-y Cr 2 O 4 with x+y≤1.

Outer oxide layer: formed by crystal agglomerates produced by the surface precipitation of ions in the primary coolant [6]. It is assumed that the surface of the inner oxide layer is coated by these crystals. In the case of nickel-and iron-based alloys for example, this oxide is a mixed oxide whose formula can be expressed as follows: Ni x Co y Fe 3-x-y O 4 with 0≤x≤1 and 0≤y≤1-x. In relation to the values of x and y, this can be magnetite, nickel ferrite, or cobalt ferrite. Metallic nickel (Ni) or nickel oxide (NiO) may also be found.

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

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

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

Filters and resins: these two systems are used to purify the primary system to remove the particles and ions. 3

Chemical elements and isotopes

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

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

• 58 Ni (n,p) 58 Co • 59 Co (n, γ) 60 Co The mass balances are performed for each isotope of these chemical elements.

Mass balance and transfer mechanisms

The behaviour of the corrosion products in each control volume is modelled by the mass balance equation for each isotope in each medium [START_REF] Marchetto | PACTOLE V3 : a new code version to predict corrosion product contamination[END_REF]: Corrosion of the base metal leads to the formation of an inner oxide layer on the one hand, and to the supply of ions in the primary coolant (net releases) and the formation of the outer oxide layer on the other hand. The formation of this outer layer is partially due to the ions coming from the base metal that precipitate after mixing with the ions in the boundary layer. Therefore, radioactive ions in the coolant can be incorporated into the outer oxide layer even in a saturated medium. This modelling is based on a pragmatic approach [8] for which the parameters -especially corrosion and release ratesare determined on the basis of test loop experiments.

( ) ∑ ∑ - = - + Source Sink m m in out i J J m m t m & & ∂ ∂ With m i
The dissolution of the deposit or the outer oxide layer can occur when the concentration of ions in the coolant is below the equilibrium concentration. The dissolution phenomenon is controlled by the transport kinetics of ions on the wall into the bulk of the coolant (mass transfer coefficient), by the oxide transformation kinetics, and by the concentration gradient.

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

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

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

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

-Turbulent diffusion and the effects of inertia [10], -Sedimentation for horizontal ducts [11], -Thermophoresis for temperature gradients between the coolant and the wall [12]. The erosion term involves the erosion of particles composing the deposit resulting from the coolant friction forces. The PACTOLE code model assumes that the deposit can be eroded above the laminar sub-layer.

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

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

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

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

SIMULATIONS USING PACTOLE V3.2

Main input data

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

-The geometric parameters of a control volume are the hydraulic diameter and its wet surface. -The thermohydraulic data such as velocity, flow and temperature of the coolant must be indicated, as well as the wall temperature. -The materials are defined by their density, their chemical composition and their roughness. Any possible surface treatment is taken into account via the experimental corrosion and release rates. -As the activation of the corrosion products is directly proportional to the neutron flux, the power fraction is specified in each part of the core. -In terms of the operating data, a cycle is defined by its duration, its power history and its concentration levels in boron, lithium, hydrogen and oxygen.

Simulation of CRUAS-1

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

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

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

The simulated primary coolant chemistry is that of the first seven cycles of CRUAS-1. CRUAS-1 operated according to a coordinated boron-lithium chemistry with a pH 300°C of 7.0 up to the third cycle, then according to a coordinated boron-lithium chemistry with a pH 300°C of 7.2. The operating parameters are given in Figure 4.

Figure 5, Figure 6 and Figure 7 compare the deposited activities for the two main radioelements ( 58 Co et 60 Co) calculated using PACTOLE V.3.2 with the deposited activities of the EMECC measurements [13] for the hot legs (HL), the crossover legs and the steam generator tubes (SG) (cold side). In relation to the 58 Co activity measurements, the CRUAS-1 simulation overestimates the contamination in the hot legs by a factor of 2 to 3, in the crossover legs by a factor of 1.5 to 5, and in the steam generators by a factor below 1.7.

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

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

-The deposited 58 Co activity tends to decrease in the legs with each cycle, and remains relatively constant in the steam generators. -The deposited 60 Co activity tends to increase for several cycles. Figure 8 The difference between the calculated figures and the experimental measurements can therefore be significant but is greatly reduced in relation to previous versions of the code, and new improvements are expected for the future versions.

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

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

Influence of an operating parameter

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

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

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

Figure 9 shows the mass of nickel deposited on the fuel that is removed by the PACTOLE code during cold shutdowns with oxygenation. It seems that the closer the unscheduled shutdown with oxygenation is to the end of the cycle, the less nickel will be removed during the end-of-cycle shutdown, and thus the weaker 58 Co peak will be.

Mass of nickel removed during shutdowns

Therefore, the simulations show that an unscheduled shutdown occurring 30 days before the end of the cycle reduces the mass of nickel removed by 75%, by 60% when occurring 90 days before, by 30% when occurring 180 days before, and by 15% when occurring 250 days before. By comparing these figures with feedback data [15], it was observed that the two trend lines are parallel and very similar (see Figure 10). Modelling the behaviour of corrosion products in a PWR primary system remains essential in terms of both present and future industrial stakes: radiation protection, reactor availability, environmental impact and dismantling.

The latest version of the PACTOLE code, PACTOLE V3.2, is an important step forward towards predicting PWR contamination. The main improvement is its coupling with a chemical module associated with a thermodynamic database that can be used to describe the speciation of the aqueous medium in interaction with several solid (metals, simple oxides, and mixed oxides) and gaseous phases in reactor primary systems. Its object-oriented language makes it possible to upgrade the modelling by making relatively easy changes to equations expressing the physical conditions in question.

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

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

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

Figure 1 :

 1 Figure 1 : Representation of a PWR primary system and a CVCS in terms of control volumes

Figure 2 :

 2 Figure 2 : Media taken into account in PACTOLE code modelling

  representing the mass of the isotope (i) in a given medium, and t representing the time, and J m representing the material flux between 2 media (except for nuclear reactions that represent a mass transfer inside the same medium). The different transfer mechanisms taken into account in the PACTOLE V3.2 mass balance equations are indicated in Figure3.

Figure 3 :

 3 Figure 3 : Mass transfer mechanisms integrated into the PACTOLE V3.2 code

Figure 4 :

 4 Figure 4 : CRUAS-1 -Operating parameters

Figure 5 :

 5 Figure 5 : CRUAS-1 -Activity deposited in the hot legs

Figure 6 :

 6 Figure 6 : CRUAS-1 -Activity deposited in the crossover legs

  compares the volume activities of 58 Co and 60 Co calculated with the PACTOLE V3.2 code with the volume activities measured by EDF. The calculated volume activities are about several MBq/t for 58 Co and several tenths of MBq/t for 60 Co. The measured volume activities are about a dozen MBq/t for 58 Co and several MBq/t for 60 Co. The calculations therefore underestimate the volume activity of 58 Co and 60 Co by about a factor of 10 or less.

Figure 8 :

 8 Figure 8 : CRUAS-1 -Volume activity levels

Figure 9 :

 9 Figure 9 : Effect of an unscheduled shutdown Mass of nickel removed during shutdowns

  unscheduled shutdown with oxygenation on the58 Co or Ni peaks during the refuelling shutdown

Figure 10 :

 10 Figure 10 : Effect of an unscheduled shutdown Comparison between PACTOLE and feedback 5. CONCLUSION

activated corrosion products with the PACTOLE V3.0 code", Proceedings of the International conference on Water chemistry in nuclear reactors systems, 11-14 October 2004, San Francisco (US), Poster ½.

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

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

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

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

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

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

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

[13] R.