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[en] In order to fulfil new superior cladding the requirements for Generation IV reactors for fission the stainless steels is improved by oxide dispersion strengthening. The austenitic 304L stainless steel was mechanical improved by addition of two nano-oxides types: titanium and yttrium oxides. The material resulted, 304ODS, was characterised by metallographic techniques and the oxide particles distribution were put in evidence by scanning electron microscopy. For the compatibility study with media from Supercritical Water Reactor, corrosion tests have been performed in supercritical water environment at a temperature of about 823K and 25MPa pressure. The weight gain was measured for 304ODS samples and compared with 304L stainless steel coupons. The oxides morphologies and stratification were examined with the aim of establish the protective character of oxides. For 304ODS samples the weight gains are higher than for the austenitic steel when the exposure time was up to 1320 hours. The oxide formed on the 304ODS samples in SCW are layered, a more uniform oxide was observed on 304ODS steel after corrosion than on 304L steel. The morphology of oxide surface was observed by SEM. The composition of sub-layers was investigated using EDX spectrometer and by Gracing Angle X-ray Diffraction. (author).
[en] Supercritical water reactor (SCWR) is one of the Generation IV reactors characterized by operation above the thermodynamic critical point of water. A preliminary design of an SCWR revealed an operating pressure of 25 MPa and a coolant temperature of up to 620 "oC at the core outlet. Operating above the thermodynamic critical point of water the SCWR offers many advantages, such as simplified design, smaller volume, and higher thermal efficiency compared to the current light water reactor (LWR). However, supercritical water (SCW) can be a very aggressive oxidizing environment, especially at higher dissolved oxygen contents, and oxidation rates in such service condition are significantly enhanced. In a highly oxidizing and high temperature environment, corrosion of component alloys is expected to be more serious than that in a typical boiling water reactor environment. It would be a quite challenging to maintain structural integrity in the SCWR environment,especially in the core region. The most promising structural materials for the SCWR are austenitic stainless steels and nickel-base alloys. Literature review showed that the most probable fuel cladding material may have an austenitic structure and contain high Cr concentration up to 22% or higher. Therefore, a systematic study on the corrosion behavior of structural materials is needed to ensure their safe application to nuclear reactor systems. This paper present a part of the research program performed at RATEN ICN on oxidation in supercritical conditions of commercially available austenitic alloys.
[en] Interaction between fuel element cladding and water coolant plays an important role in normal operation, can have a dominant role in accidental situations and can lead to failure of fuel rods and activity release. For the future, the tendency will be to increase the coolant temperature, extend fuel residence time in the reactor core (for higher burnup) and increase the heat flux. This can lead to increased probability of fuel failures due to waterside corrosion, corrosion products accumulation and deposition. In order to prevent cladding failures, the coolant chemistry must be monitored and controlled in order to reduce the amount of deposited crud and the oxygen potential. Corrosive deposits together with aqueous corrosion influence the performance of fuel elements by increase of temperature on cladding surface or changes in the coolant chemistry (increase of water pH), phenomena which lead to cladding failures. The process of corrosion products formation on zircaloy-4 fuel cladding surface and their consequences was evidenced by performing of experiments in: autoclaves circuits assembled in a by-pass loop of a CANDU-6 Reactor at NPP Cernavoda; irradiation loop of the TRIGA Reactor, and in laboratory static autoclaves. The determination of corrosion and the characterization of crud deposits on the zircaloy-4 surfaces were performed using gravimetric method, metallographic and electronic microscopy, and gamma spectrometry analysis and impedance electrochemical spectroscopy (EIS) determinations. The experimental results showed that the composition, thickness and evolution of corrosive deposits on fuel assembly surfaces depend very much on operational conditions, such as steady state operation, water chemistry conditions (pH and oxygen concentration) and different oxidation conditions of cladding surface. (author)
[en] Development of this system implies investigation of the structural materials corrosion processes, in different conditions of water chemistry and temperature, for understanding of the corrosion degradation phenomena that conduct to failure of some components from PHTS of CANDU reactor. In this aim, a chemical control of the water and structural materials corrosion testing programme, as well as a system for data acquisition and processing, were established. On the basis of the experimental data, obtained from in-pile and out-of-pile corrosion experiments, on different structural materials, a programme on computer was developed, for analysis of their corrosion behavior. (author)
[en] The study is focused on the hydrogen content measurement method for the Zr-2.5%wt Nb CANDU reactors pressure tube samples, by means of the image analysis techniques. The metallographic pictures of the controlled hydrogen samples have been processed as digital images. A Visual Basic software program was developed in order to automatically process digital images acquisitioned by an optical microscope in order to evaluate the hydrogen contents. A relationship between the area fraction and the hydrogen content was established. The method can be applied for samples with hydrogen content in a range of 20 - 200 ppm. The proposed method can be seen as an alternative hydrogen measurements method to the fusion method. (author)
[en] Pitting corrosion is potentially damaging form of corrosion encountered in metallic components in the cooling water systems of CANDU steam generator. This paper presents some methods which can quantify localized corrosion. There are many different electrochemical techniques used to determine characteristic pitting potentials. From these methods, potentiodynamic cyclic polarization is most widely used as a predictive tool for characterizing the tendency of a material to experience pitting corrosion. Ideally, the pitting scan would be performed in the same solution to which the metal would be exposed in service. The purpose of the present investigation was to establish the corrosion behaviour of CANDU steam generator components in normal and abnormal operation conditions of secondary steam generator circuit. The corrosion of alloys is caused by local breakdown of the protective oxide films when aggressive chemical environment accumulates in restrictive flow areas from secondary side of steam generator. Because the chemical composition and the structure of oxide layers play a prominent role in corrosion process, characterization studies of these oxide layers is essential to perform. The paper presents laboratory corrosion studies which have been performed on the steam generator shell material (carbon steel SA 516) and the material of tubing (Iy 800). The samples used in tests were oxidized in static autoclaves, under normal water chemistry of steam generator secondary side. The susceptibility at pitting corrosion of the oxidized and non-oxidized carbon steel samples and Iy 800 samples were investigated by cyclic polarization in simulated secondary steam generator water solutions with different NaCl concentrations. In addition, monitoring of the open circuit potential of the samples gave information about the metal/electrolyte reactions. Characteristics of the oxide films were studied by optical microscopy and scanning electron microscopy (SEM)
[en] The zirconium alloys oxidation and the oxidation kinetics is governed by the diffusion of oxygen in the oxide scale formed at the surface of the metal, and the oxide scale is formed normally under compression. With the increase of the oxide scale thickness the compressive stress at the oxide-metal interface increase until plastic deformation of the oxide and metal ensues, meanwhile the compressive stress at oxide-environment interface decreases and may become tensile. The kinetic curve reflects the changes of the compressive stress by changes in the rate of oxygen uptake. The volume expansion associated with this transformation is the cause of the oxide cracking. The crack initiation is the primary cause of the cyclicity of the post-transition kinetic curves at temperatures up to 700°C. The purpose of the present paper is to correlate the oxide scale thickness at cracking obtained from the kinetic curves and those obtained by SEM. (authors)
[en] The reliability of all nuclear power plants being dependent on the water chemistry during normal operation, start-ups, shutdown and abnormal operation cases, the water chemistry control is very important. It is necessary mainly to maintain those conditions that should assure minimum materials corrosion and the components reliability in operation. The corrosion may be minimized both by right choice of materials and by narrow chemical control of aqueous environment in contact with metallic materials. Referring to water chemistry management, it found that some amines simultaneously with their action of pH regulating, should have an inhibitor effect on the corrosion behaviour of iron alloys. The alloy studied was the carbon steel SA106 gr.B. The carbon steel samples were submitted to autoclavizings at (260 ± 5)°C and (5.5 ± 0.5) MPa in the three types of solutions having a similar final pH of 9.7, prepared using the following amines: morpholine and cyclohexylamine (AVT), morpholine and triethanolamine (TEA) and respectively morpholine and triethylamine (TREA). After autoclavizings, the filmed samples have been characterized using the gravimetric method, Electrochemical Impedance Spectroscopy (EIS), SEM and metallographic evaluation. On the basis of our experimental results, we concluded that triethanolamine and triethylamine may successfully substitute cyclohexilamine in the chemical treatment of secondary circuit environment. (author).
[en] In order to fulfil superior cladding for new reactor generation G IV, the austenitic""3"0"4L stainless steel was improved by oxide dispersion strengthening (ODS), using two nano-oxides: titanium and yttrium oxides. The behaviour of the new material resulted, 304 ODS, in water at supercritical temperature of about 550"OC and 25 MPa pressure, was considered. The oxidation kinetics by weigh gain measurements for both materials have been estimated and compared. The weight gain of ODS samples is higher than basic austenitic steel up to 1320 hours. The oxides developed on the ODS samples in SCPW are layered and more uniform than in "3"0"4L SS. The protectively character of oxide films was estimated by different techniques. The morphology of oxide surface, the layering and chemical formula of oxides films were investigated by scanning electron microscopy (SEM), Energy Dispersion X-Ray Spectrometry (EDS), electrochemical impedance spectrometry (EIS) and by Small Angle X-ray Diffraction (SAXD). 1. (authors)
[en] A nuclear power reactor is a very complex system, with many parameters influencing the oxidation and hydrogen absorption kinetics of zirconium alloys, used as structural materials for reactor fuel components in light water or heavy water cooled nuclear reactors. Contamination of the cooling water is a most undesirable situation, so it is necessary that the fuel rods remain leak free through the life of the fuel. The pH-value is one of the important parameters for the oxidation behavior of Zirconium alloys inside a reactor. To study the influence of coolant pH on Zircaloy-4 fuel claddings oxidation and hydriding, in operation condition with high burn-up fuel, the corrosion experiments in LiOH solutions with several concentrations, at high temperature and pressure (310 deg. C and 10MPa) will be performed, on initial non-oxidized and different oxidized samples. (author)