19 research outputs found
Activated corrosion product contamination assessments of DEMO WCLL breeding blanket primary heat transport system
No full textIn water-cooled fusion reactors, the assessment of the primary system contamination is essential for waste management, machine availability, occupational radiation exposure, and radiological hazard determination. The primary cooling water is not only directly activated by the intense neutron field but is a contamination vector for a significant variety of gamma emitters with short to long decay half-lives. Corrosion products can be activated in those regions under neutron flux of the primary circuit and then released in the cooling water.
In the EU-DEMO fusion power plant equipped with the Water-Cooled Lithium Lead Breeding Blanket (WCLL-BB) concept, the primary coolant undergoes intense neutron fields in the first wall and the breeding zone regions of the blanket. Activated Corrosion Products (ACPs) are then formed, released into the water, transported in the cooling loop and finally deposited onto the ex -vessel surfaces of the Primary Heat Transport System (PHTS), where working personnel are susceptible to being radiologically exposed.
This work addresses the complete assessment of ACPs in the WCLL-BB PHTS of EU-DEMO. The simultaneous and multi-physical processes behind the ACP formation are tackled using the OSCAR-Fusion code, a comprehensive tool developed by the CEA (France) to assess contamination in fusion nuclear reactors. The whole system is modelled with zero-dimensional nodes with assigned geometrical, thermal-hydraulics, material and chemical parameters. Activation reaction rates integrated over the whole spectrum and calculated with MCNP are given to those regions exposed to the neutron flux. Results are provided in terms of mass and activity inventories of ACPs as deposit and inner oxide layers of components (pipes, heat exchangers, pumps...), ions in solution, particles in suspension, and filters and resins trapping.
Mobilizable inventories such as ions, particles and deposits are important source terms in accidental scenario evolutions, while the whole activity inventory constitutes the main long-term gamma emitting source for dose rate maps determination in the tokamak building rooms housing the main PHTS equipment
Modelling of the contamination transfer in nuclear reactors: The OSCAR code - Applications to SFR and ITER
No full textInternational audiencePredicting the radioactive contamination of nuclear reactor circuits is a significant challenge for plant designers and operators. The major stakes are to decrease personnel exposure to radiations, to optimize plant operation, to limit activity of wastes and to prepare decommissioning. To address this challenge, the French strategy has been focusing on performing experiments in test loops representative of Pressurized Water Reactor (PWR) conditions, measuring the PWR contamination and developing a simulation code known as OSCAR. The OSCAR code has been developed by the CEA in collaboration with EDF and AREVA NP for more than 40 years. It has resulted from the merging of two former codes PACTOLE for Activated Corrosion Products (ACPs) and PROFIP for fission products and actinides. Thus, the OSCAR code has been designed to simulate the contamination transfer in the PWR Reactor Coolant System (RCS). The main transfer processes involved for ACPs are corrosion-release, erosion, deposition, dissolution/precipitation, convection, purification, neutron activation and radioactive decay. The OSCAR code calculates the space-time distribution of activities in solid and liquid phases in the PWR RCS and auxiliary systems. It is a modular code and it can easily integrate new models.In the same way, ACPs in the primary heat transfer system of the International Thermonuclear Experimental Reactor (ITER) and in the primary system of Sodium Fast Reactors (SFRs) can be a major concern as contributors to the source term of potential released activities to the environment in case of accident and to the occupational radiological exposure during the normal operation. Although the design and operating conditions of these reactors are different from those of PWRs, an application of the OSCAR code to SFRs, so-called OSCAR-Na, and another one to ITER, so-called OSCAR-Fusion, have been developed to treat the behaviour of corrosion products thanks to the modularity of the OSCAR code.After a general presentation of the OSCAR code, this contribution will describe the modelling of the ACP transfer common to the different types of nuclear reactors (discretization into control volumes and media, metallic elements, transfer mechanisms, mass balance equations) and then the specificities of OSCAR-Fusion and OSCAR-Na (in particular the corrosion model of OSCAR-Na). Finally, issues and RetD needs will be presented
Validation of European computer codes used for fusion safety analysis
No full textThe development of safety files for fusion plants rely on several calculations and safety codes used for computing various situations and events. Within the future safety files, several safety related simulation codes are used for demonstrating the safe behaviour of the concerned machine and the limited impact on the population and environment in various situations from the normal operation up to the largest credible accident. These codes are ranging from neutronics calculation and dose assessment to the modelling of accident sequences and thermo hydraulics analysis in normal, incidental and accidental situations.
An important aspect for the safety analysis process is the validation of the models and the reliability of the calculation results. This necessitates simulating in actual facilities physico-chemical situations related to the specific validation needs of the concerned codes.
Several of these computer codes were already applied for long time in the fission industry and have been validated for this purpose. Fusion, however, requires the development of specific codes or the extension of existing ones to deal with all fusion specific conditions. These newly developed or extended computer codes have to be validated by conducting representative experiments or by cross checking various codes based on different modelling approaches .
The paper reviews the most important computer codes applied for fusion plants and ITER safety analysis together with the validation efforts carried out in the various European Member States. Experimental data and facilities used for this purpose will also be presented. Finally, the paper addresses the still remaining gaps in the code validation and provides guidance on needs for further code development
Corrosion of 316L exposed to highly concentrated borated water used as shield in nuclear fusion experimental reactors cooling circuits
No full textBorated water is used as a shield in nuclear fusion cooling circuits. General corrosion, activated corrosion products (ACPs) formation and stress corrosion cracking initiation of 316 L steels exposed to ultrapure water (UPW) or 8000 ppm B water at 80°C were tested. A Ni enriched sub-oxide layer, a transition layer and oxide layer were observed using advanced characterisation (STEM-EELS, APT). The oxide formed in UPW was pro tective (Cr:O 40:45), the oxide formed in 8000 ppm B was non-passivating. 8000 ppm B led to higher release of Fe, Cr and Mo, 316 L was more prone to SCC initiation and enhanced ACPs formatio
Influence of the dissolved hydrogen concentration on the contamination of the primary loop of DOEL PWR using the OSCAR code
No full textInternational audienceNormal operation of PWR generates corrosion products in the primary circuit which are activated in the core and constitute the major source of the radiation field. To minimize these risks and constraints, it is essential to understand the behavior of corrosion products by carrying out the appropriate measurements in PWR circuits and loop experiments combined with reactor contamination assessment code. The reactor contamination assessment code allows to take into account the chemical and physical mechanisms in different situations in operating reactors or at design stage. OSCAR code has been developed with this aim by CEA in collaboration with EDF and Framatome, and has actually been used since the early seventies [1]. OSCAR is a reliable tool for PWR (also used for EPR, SFR, ITER, decommissioning…) calibrated and validated with a complete database of contamination measurements in EDF water reactors. Water chemistry has an influence on corrosion of the main materials (especially nickel-based alloys), in Belgian PWR the average dihydrogen concentration used is around 30 cc/kg, which is in line with current international standards however which is not the best value to reduce stress corrosion cracking of the materials. It also has an influence on dissolution/precipitation mechanisms involved in contamination. Water chemistry control allows then to reduce significantly the radioactive contamination in the primary loop and therefore facilitates maintenance operations. In this field, hydrogen plays a critical role in limiting the presence of oxidizing species due to water radiolysis. Increasing hydrogen could also reduce core internals cracking. The aim of this study is to evaluate the influence of an increase/decrease of the Dissolved Hydrogen (DH) concentration on the contamination of the primary loop using the OSCAR code. This study presents the simulation results of a sensitivity analysis, using the 1.3 version of the OSCAR code, of the contamination in the primary loop of DOEL PWR with DH concentrations ranging between 15 and 70 mL/kg.The evolution of the surface contamination in Co and Co were calculated on the hot legs, U tubes and Steam Generators (SG) tubing along with the mass of Ni deposited on the fuel. In order to explain those evolutions, equilibrium (concentration of the element in the coolant with respect to the considered oxide) and ionic Ni concentrations were investigated on the SG and the fuel as well as the Ni dissolution and release flux of the SG
Zinc effect on the primary circuit contamination of a Belgian PWR using the OSCAR V1.3 code
No full textInternational audienceThe purpose of the OSCAR code, developed by the CEA in collaboration with EDF and AREVA NP, is to simulate the contamination of the PWR (Pressurized Water Reactor) primary system. The OSCAR code, coupled to its chemistry module PHREEQCEA, allows researchers to analyse the corrosion product behaviour and to calculate the volume and surface activities of a primary system. It synthesizes knowledge and allows us to progress in understanding the phenomena involved in the contamination caused by activated corrosion products.The OSCAR V1.3 code has been used to simulate the effect of zinc on the primary circuit contamination of a Belgian PWR operated by ENGIE. This paper highlights the physicochemical phenomena involved during a zinc injection. The OSCAR simulation shows thatZinc precipitates on the inner oxide (Cr-rich layer) instead of cobalt showing that zinc has a better affinity than cobalt with chromites (change in the Co equilibrium concentration).Cobalt concentration and Co and Co activities increase in the primary coolant and on the outer oxide of both the in-core and out-of-core surfaces (Fe-rich layer). Different plants have observed this increase during the first cycles of Zn injection.An antagonist effect exists between the decreasing Co and Co activities of the inner oxide and the increasing activities of the outer oxide, which explains the slight impact of a zinc injection observed by EMECC campaigns
The OSCAR V1.4 code: A new dissolution-precipitation model to better simulate the corrosion product transfer in nuclear cooling systems
No full textInternational audiencePredicting the radioactive contamination of nuclear reactor circuits is a significant challenge for plant designers and operators. To address this challenge, the French strategy has been focusing on performing experiments in test loops, measuring the PWR contamination and developing a simulation code so named OSCAR.The process governing the contamination by Activated Corrosion Products (ACPs) of a nuclear cooling system involves many different mechanisms that react with each other. One of the most important mechanisms is the dissolution-precipitation mechanism, which governs the behavior of soluble corrosion products and which is related to water chemistry specifications.The dissolution-precipitation model has been improved in the new version of OSCAR, OSCAR V1.4. Thanks to this improvement and to the OSCAR chemistry module, PHREEQCEA, the OSCAR V1.4 code can reproduce the impacts of pH and of Zn injection on Co contamination highlighted in a laboratory experiment and can better reproduced the volume activity variations during a cold shutdown.This new version of the OSCAR code is a powerful tool to predict the contamination of nuclear systems and to analyze the corrosion product behaviors in different conditions and thus to provide explanations of these behaviors
The OSCAR V1.4 code: A new dissolution-precipitation model to better simulate the corrosion product transfer in nuclear cooling systems
No full textInternational audiencePredicting the radioactive contamination of nuclear reactor circuits is a significant challenge for plant designers and operators. To address this challenge, the French strategy has been focusing on performing experiments in test loops, measuring the PWR contamination and developing a simulation code so named OSCAR.The process governing the contamination by Activated Corrosion Products (ACPs) of a nuclear cooling system involves many different mechanisms that react with each other. One of the most important mechanisms is the dissolution-precipitation mechanism, which governs the behavior of soluble corrosion products and which is related to water chemistry specifications.The dissolution-precipitation model has been improved in the new version of OSCAR, OSCAR V1.4. Thanks to this improvement and to the OSCAR chemistry module, PHREEQCEA, the OSCAR V1.4 code can reproduce the impacts of pH and of Zn injection on Co contamination highlighted in a laboratory experiment and can better reproduced the volume activity variations during a cold shutdown.This new version of the OSCAR code is a powerful tool to predict the contamination of nuclear systems and to analyze the corrosion product behaviors in different conditions and thus to provide explanations of these behaviors
mAg behaviour in PWRs Lessons learnt from the EMECC campaigns
No full textInternational audienceFor 45 years, most of the time in collaboration with EDF, CEA has measured the contamination of PWR circuits by gamma spectrometry using the so-called EMECC device. These measurements allow to determine the surface activities accurately in order to study the behaviour of the activated corrosion products.In case of pollution by mAg, the main lessons learnt from EMECC campaigns and primary coolant filtering campaigns as well are as followsDuring oxygenation, Ag dissolves from the primary surfaces and precipitates on the cold parts of auxiliary systems before dissolving slowly from these cold parts. It leads to a sharp increase of dose rate at the vicinity of the cold parts of auxiliary systems.Because of the precipitation on the Nuclear Sampling System (NSS), 110mAg sampled via the NSS is not representative of the Ag volume activity of the Reactor Coolant System.Under oxidizing and acid conditions, Ag is trapped very well by cation exchange resins. On the other hand, silver is in an insoluble form (colloids) in a reducing medium and it is not trapped by ion exchangers
Development of the PACTITER code and its application to safety analyses of ITER Primary Cooling Water System
No full textInternational audienceThe PACTITER code derives from the PACTOLE code, developed by the CEA for predicting Activated Corrosion Products (ACPs) in PWR primary circuits. The operating conditions, material compositions and water chemistry of the various Primary Heat Transfer Systems (PHTS) of the International Thermonuclear Experimental Reactor (ITER) made mandatory the adaptation of the PACTOLE code. This new code, called PACTITER, has then been developed on the basis of dedicated experiments, and particularly, in order to determine the stainless steel release, which depends on the thermal-hydraulic conditions (fluid temperature and Reynolds number). The PACTITER code has been used in support of the ITER Generic Site Safety Report (GSSR) in the field of accident analysis and worker collective dose assessment
