Modeling particle deposition in the primary circuit of pressurized water reactors for the OSCAR code

International audienceParticulate corrosion products are generated in the primary system of pressurized water reactors (PWRs) by volume precipitation and by erosion of oxides formed on metal surfaces through their uniform corrosion. The activation of corrosion products, mainly 58Co and 60Co (respectively coming from the activation of 58Ni and 59Co) leads to radiation field growth around the primary system, directly impacting system integrity and the radioprotection of nuclear workers. In order to understand and mitigate contamination by activated corrosion products, contamination predictions can be performed using the OSCAR code, which relies on the development of models to describe the numerous and complex interactions at stake. Particulate corrosion products account for a significant portion of corrosion products, as such the deposition/erosion mechanisms have their importance for the overall computation of surface or volume contamination. The aim of this article is to present an updated and inclusive deposition model for particulate corrosion products by taking into account surface interactions. The impact of the new deposition model on contamination predictions is then evaluated and has enabled to reproduce, for the first time using the OSCAR code, the preferential contamination in 58Co in the cold side of the circuit, measured by gamma spectrometry with the EMECC device on commercially PWRs

Long term trend of the cobalt activity in the primary coolant of Beznau NPP evaluation of the operational and shutdown chemistry

International audienceThe nuclear power plant Beznau consists of two pressurized water reactors which went into operation in 1969 resp. in 1971. Numerous back-fittings were carried out like the replacements of the steam generators (SGR) realized at unit 1 in 1993 and at unit 2 in 1999. Since the SGR special attention has been paid to the behavior and inventory of cobalt nuclides and their contribution to the contamination level which is approximately 98% of the dose rate. A lot of different measures have been realized to decrease the dose rates. The results were checked by different investigations such as in situ-spectrometry and analyses of the corrosion products in the coolant. A shutdown chemistry was introduced in 1982 to stop the increasing trend of the dose rates, to have definite redox conditions, to remove the released activity, to realize a rapid degasification, to reach an activity in the pool water as low as possible and a high visibility of the water. During shutdown corrosion products are mobilized as a result of changes in power, temperature and water chemistry. This process is supported by the dosage of hydrogen peroxide. The peak of 58Co is 10 to 15 GBq/m 3 whereas about 4000 GBq 58Co. The usual effect of the SGR on the increase of the 58Co activity cannot be avoided because the accommodation of the new 690 TT surfaces during the subsequent cycles can temporary lead to higher nickel release compared to "stabilized" older 600 alloy. Now the increase of cobalt nuclides seems to have been enhanced by additional and simultaneous conditions: higher nickel and 58Co contents at the beginning of cycle after the short outage of cycle 26 and change in the primary pH 300 value from 7.2 to 7.4 accompanied by a higher upper lithium limit of 2.5 ppm. Especially that may enlarge the formation of metallic nickel which is the main source of coolant-born 58 Co. This may cause an increased crud release which also affects the reactor core zones where precipitation of Ni (m) increases the oxide film heterogeneity and instability of the layer as a whole. Solubility calculation results show that the driving forces for the deposition of Nickel species in the core are the lowest solubility limit and negative solubility coefficients. The rather high hydrogen content in the RCS at Beznau units imposes that metallic Nickel with a low solubility value (about 0.05 ppb) and a slightly negative coefficient is always present in the core. Moreover its solubility limit is lower for higher pH 300 values, that means between 7.0 (BOC) and 7.2 or 7.4 (target value). It is to note that the last change of the coolant chemistry does not meet the optimum operating conditions for the Beznau units. The increase of the 58 Co activity seemed to have been higher than normally expected after a SGR with alloy 690 TT tubing material. Therefore a decision was made to return to the former target pH 300 value of 7.2

Characterization of contamination: french analysis of international EMECC campaigns

International audienceThe reduction in collective dose is an important objective for all nuclear plant operators. One of the ways for studying the impact of design and operation parameters is the measurement campaigns of contamination in PWRs. EDF' strategy is based on several levels of measurement: local dose rates give raw information; dose rate index and CZT spectrometry allow an evaluation of the radiological situation of each plant, its evolution and comparison with other plants. For very accurate characterization, EDF performs gamma spectrometry measurements with CEA and its EMECC system, with more than 300 campaigns in EDF plants and also about 70 campaigns in collaboration with other operators. Recent collaboration between EDF and other operators with excellent collective dosimetry resulted in measurements during several Refuelling Outages. The present paper presents results obtained with EDF Energy at Sizewell B Outage 8, ELECTRABEL at Doel 3 Outage 25 and Doel 4 Outage 22 and CNAT at Trillo Outage 21. The paper presents the results in terms of deposited activities and dose rates in RCS and CVCS, and compares them with French PWRs, and especially with Saint-Laurent B1. The specificities of all these plants are identified, to explain the differences in results, either for dose rates, deposited activities or collective dosimetry. All these comparisons will enable EDF and collaborative operators to improve the performance of their own plants

Development of PACTITER code and its application to safety analyses of ITER Primary Cooling Water System

International audienceThe PACTITER code derives from the PACTOLE code, developed by the CEA for predicting Activated Corrosion Products (ACPs) in PWR primary circuits. The PACTOLE code was initially used (1994-1997) for predicting ACPs in the various stainless steel based Primary Heat Transfer System (PHTS) loop of the International Thermonuclear Experimental Reactor (ITER). The operating conditions, material compositions and water chemistry of ITER PHTS (no soluble Boron, lower temperature range (100-240°C), and mainly the presence of Cualloy in the divertor cooling loop) made mandatory the development of a dedicated code. Required chemical data were provided by "ad hoc" experimental tests, carried out at CEA Fontenay-aux-Rose (1997) and aimed to derive Cu solubility law in ITER relevant conditions. The application of such new code, named PACTITER, to divertor cooling loop was promising, as the predicted Cu-alloy release rates were in agreement with experimental data. The PACTITER code was then applied to predict the ACP source term for all ITER EDA (1998) PHTS cooling loops: first-wall/blanket (10 loops), limiter/baffle (4 loops), divertor (4 loops) and vacuum vessel (2 loops). The ACP inventory and its distribution on loop components, predicted by the code, were used as input for the accident analyses related to LOCA and to foresee the collective dose to the staff involved in scheduled and unscheduled maintenance. Related results were documented in the ITER NSSR-2 (Non-Site Specific Safety Report 2). Following the evolution of the ITER design (2001) to a reduced-cost machine, the PHTS was completely redesigned with a drastic reduction of number of loops: 3 for the primary first wall/blanket, 1 for the divertor/limiter and 2 for the vacuum vessel. The PACTITER code was again used in support of accident analysis and worker collective dose assessment, with results documented in the Generic Site Safety Report (GSSR) (2001). Pending the final decision for the ITER site, in order to improve the validation level of PACTITER, tests were performed by CEA in 2001 and 2004 in the CORELE test facility. The comparison between experimental results and PACTITER predictions has provided good results, revealing the influence of the Reynolds on the material release rate. A new version of PACTOLE (version 3) has been designed to account for recent development relevant to the behaviour of corrosion products: improvement of the corrosion/release model, of the oxide and equilibrium concentration determination. Moreover, the data processing architecture and the numerical analysis for solving transient equations have been totally reviewed. It is foreseen that the PACTITER code will profit by these improvements

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

International audienceThe reduction of collective dose is an important objective for all nuclear plant operators. One of the support for studying the impact of design and operation parameters is the measurement campaigns of contamination in PWRs. EDF performs consequently gamma spectrometry measurements with CEA and his EMECC system, with more than 200 campaigns in EDF plants and also about 70 campaigns in collaboration with other operators. A recent collaboration between EDF and British Energy resulted in measurements during Refueling Outage 8 at Sizewell B, a PWR with excellent collective dosimetry. The paper presents the results in term of deposited activities and dose rates in RCS and CVCS, and compares with French PWRs, and especially with Saint-Laurent B1. If the dose rates are very comparable, the collective dosimetry is slightly lower for Sizewell B. First explanations are searched for the established differences. Lower 58 Co activity for Sizewell B is probably due to the good manufacturing of alloy 690 steam generator tubing, higher 60 Co activity is probably linked with the use of alloy 718 grids, not totally removed after 8 cycles. The origin of difference in collective dosimetry should be less time spent in dosing outage operations, even if the total time spent in controlled area is rather higher in Sizewell B. An accurate comparison for specific chosen outage operations is in progress. Another explanation should be differences in contamination on auxiliary systems, or differences in volume activities. To explore further these hypothesis, complementary measurements are planned by BE and EDF. All these comparisons will give new elements to BE and EDF to improve the behaviour of their own plants

Contamination simulation of DOEL-4 PWR using The OSCAR V1.2 code and parametrical studies during cycles or cold shutdowns

International audienceThe purpose of the OSCAR code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the contamination of the PWR primary system. The OSCAR code allows researchers to analyze corrosion products and fission products behavior and to calculate the activity in the fluid and the surface activity in the primary system. Nowadays, the OSCAR 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 (ACPs). The OSCAR V1.2 code is used for ACPs simulation studies of DOEL-4 reactor. The 22 first cycles have been simulated and compared with measurements [1] of surface activities on the loops. Since, this simulation is the reference of DOEL-4 parametrical studies done with OSCAR V1.2. The effect on contamination by ACPs of a 4-month operation at 50% power during a cycle will be introduced. The results will be focused on volume and surface activities during the cycle and the following shutdown. Then several studies on mid-cycle shutdowns of a given length of time or with startup criteria on 58 Co volume activity will be presented. Last, results of parametrical studies performed on early oxygenation during cold shutdown will be given

Chemical degassing on EDF units - Feed back experience and method

International audienceThe chemical degassing feasibility of all French units RCS during outages is usually performed on the four 1500 MWe French units since July 2004 and is planned for 2007 on the 34 units of 900 MWe and the 20 units of 1300 MWe units. It helps to an optimized shutdown planning management, especially in case of unplanned shutdowns, and helps to limit the thermal stress corrosion of RHR at temperature higher than 120°C by avoiding temperature step extension after bubble collapsing to remove hydrogen increase. Chemical degassing is performed at temperature equal or lower than 80°C, by hydrogen peroxide injection if 3 < H2dissolved < 35 cm3/kg. H2O2 calculation can be stoechiometric or increased by 20 % depending on the SG tubes material and the full power operation duration before the shutdown. Purification flow is adapted to each design type units characteristics to easily manage a hypothetical early oxygenation. Early oxygenation management consists in forecasting the dilution of the VCT (Volume Control Tank) gaseous phase by nitrogen to avoid dangerous H2 /O2 gaseous mixture and to obtain oxygenation criteria on VCT. Three kinds of dilution can be used: nitrogen burping with high levels amplitude, continuous nitrogen flushing or discontinuous flushing with a maximal constant VCT level but pressure variation

Experimental study of particle deposition kinetics in the primary circuit of the CIRENE loop and comparison with OSCAR V1.2 code simulations

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

International audienceThe understanding of Activated Corrosion Product (ACP) transfers in the primary circuit is of great importance for the PWR industry. For this purpose, the CEA, in collaboration with EDF and AREVA NP, has launched an R&D program for many years involving specific experiments related to the development and validation of the PACTOLE code. Among them, the CIRENE out-of-pile loop is dedicated to the study of ACP deposits on fuel rods and on steam generator tubes. The first results were presented at the Jeju Island International Conference. This paper presents the experimental results of a new CIRENE test and compares the numerical simulations obtained with the PACTOLE V3.2 code. This test has been carried out with thermal-hydraulic parameters corresponding to a subcooled boiling regime at the outlet of the core section, and chemical conditions representative of a French PWR primary circuit. A radioactive tracing methodology have been applied, using gamma spectrometry measurements with the injection of specific radiotracers, 58 Co and 59 Fe: in-situ gamma spectrometry measurements are performed at the outlet of the core section and the SG tubes, while frequent radiochemical analyses of the primary fluid are carried out for checking the instantaneous mass balances. The comparison between the experimental results and simulations with the PACTOLE V3.2 code has led to the main following comments: • deposited activity variations in operation : for the core section as well as for the SG tube section, the calculated variations reproduce the experimental ones; • residual deposited activities after shutdown : the calculated activities are coherent with the experimental measurements. Consequently, the modelling developed in the PACTOLE code is globally validated for the CIRENE loop, and the differences between simulation and experimental results point out some remaining issues in the modelling of ACP behaviour in PWRs

Auxiliary system contamination in French PWRs

International audienceIn order to study the impact of design and operating parameters on French PWR contamination, the CEA, by request of EDF, has been performing in-situ gamma spectrometry measurements using the so-called EMECC device for 37 years. These measurements have allowed researchers to characterize contamination levels accurately (surface activities by radionuclide), mainly in primary systems, but also in some reactor auxiliary systems such as the Chemical and Volume Control System (CVCS), the Residual Heat Removal System (RHRS), the Nuclear Sampling System (NSS) or the Fuel Pool Cooling and Cleanup System (FPCS). Based on EMECC campaigns performed in different standardized plant series, this paper shows that the contamination of nuclear auxiliary systems presents differences in comparison to the contamination of the primary system:-The Co-58 and Co-60 deposited activities on the CVCS, RHRS, NSS and FPCS are generally lower. On the other hand, the Ag-110m contamination is higher with a considerable recontamination resulting during cold shutdowns.-The dose rates around the auxiliary systems are generally higher.-The Co-58 contribution and especially the Ag-110m contribution are both higher. For auxiliary systems that do not operate for several months, the Co-60 contribution emerges as the major one.-The nuclear auxiliary systems are the only circuits that allow us to measure the Cr-51 surface activities.These characteristics are due to different design and operating conditions, particularly the following : the presence of heat exchangers and lower fluid temperature, different flow rates, thinner thickness of pipes and of exchanger shells, the presence of filtering systems and operations during a short period. A final observation is that, the peculiar behavior of Ag-110m inside the NSS generally results in a poor estimation of the Ag-110m volume activity in the primary coolant