The Influence of Material Variables on Corrosion and Deuterium Uptake of Zr-2.5Nb Alloy During Irradiation PDF Download
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Author: VF. Urbanic Publisher: ISBN: Category : Corrosion Languages : en Pages : 27
Book Description
Current CANDU2 reactors use Zr-2.5Nb pressure tubes that are extruded at 1088 K, cold-drawn 27%, and autoclaved at 673 K for 24 h. This results in a metastable, two-phase microstructure consisting of elongated ?-Zr grains surrounded by a network of ?-Zr filaments. To develop a mathematical model of corrosion and deuterium ingress in pressure tubes, we have considered the impact of variables including: fast neutron flux, temperature, and the asfabricated microstructure and its evolution during irradiation.
Author: VF. Urbanic Publisher: ISBN: Category : Corrosion Languages : en Pages : 27
Book Description
Current CANDU2 reactors use Zr-2.5Nb pressure tubes that are extruded at 1088 K, cold-drawn 27%, and autoclaved at 673 K for 24 h. This results in a metastable, two-phase microstructure consisting of elongated ?-Zr grains surrounded by a network of ?-Zr filaments. To develop a mathematical model of corrosion and deuterium ingress in pressure tubes, we have considered the impact of variables including: fast neutron flux, temperature, and the asfabricated microstructure and its evolution during irradiation.
Author: BD. Warr Publisher: ISBN: Category : Characterization Languages : en Pages : 26
Book Description
Understanding the reasons for variability in D uptake in Canadian Deuterium Uranium (CANDU) reactor Zr-2.5Nb pressure tubes (PT)s will lead to improved surveillance and predictive strategies. Results from out-reactor aqueous exposures that suggest PT performance should be predictable based on offcut characteristics and may lead to the development of techniques to identify PTs with highest D uptake rates. The range of out-reactor corrosion and D uptake rates in Zr-2.5Nb coupons from different PT offcuts is similar to that found in test/power reactors. Out-reactor post-transition corrosion and D uptake rates at 310 and 360°C are found to be strongly correlated and decrease with increasing concentrations of Fe and Cr in the alloy. In the Bruce 3 reactor, PTs with higher D uptake show 'curly' ?-? microstructures with a large number of basal planes (in the ?-Zr grains), and ?-Zr grains, oriented towards the radial direction of the tube. Since oxidized ?-Zr regions are found to be associated with lateral cracking, the presence of a higher frequency of ?-Zr grains aligned normal to the surface in these curly microstructures may result in increased routes for D uptake in this material. Oxide structure is also found to be dependent on the ?-Zr orientation that results in different relative proportions of [001] growth versus general columnar oxide grains. Improved corrosion and D uptake performance is also found in Zr-2.5Nb material that is prefilmed (400°C for >=24 h), ?-quenched in the billet stage and in material from the back of the extrusion.
Author: VF. Urbanic Publisher: ISBN: Category : Corrosion Languages : en Pages : 17
Book Description
The in-reactor corrosion and hydrogen pickup of Zircaloy-2 and Zr-2.5Nb pressure tube materials are being studied in two test loops: a light water loop in the NRU research reactor, and a new heavy water loop in the Halden reactor. The complimentary test programs examine the corrosion behavior of small specimens as a function of fast neutron flux and fluence, temperature, water chemistry, and specimen pre-oxidation.
Author: Frederic R. Shober Publisher: ISBN: Category : Metals Languages : en Pages : 120
Book Description
The effect of fast-neutron (>1 Mev) irradiation on the mechanical properties of structural metals and alloys was studied. Although the yield strengths and ultimate tensile strengths are increased su stantially for most materials, the ductility suffers severe decreases. This report presents these changes in properties of several structural metals for a number of neutron exposures within the 1.0 x 10 to the 18th power to 5.0 x 10 to the 21st power n/sq cm range. Data summarizing these effects on several classes of materials such as carbon steels, low-alloy steels, stainless steels, Zr-base alloys, ni-base alloys, Al-base alloys, and Ta are given. Additional data which show the influence f irradiation temperatures and of post-irradiation annealing on the radiation-induced property changes are also given and discussed. Increases as great as 175% in yield strength, 100% in ultimate strength, and decreases of 80% in total elongation are reported for fast-neutron exposures as great as 5 10 to the 21st power n/sq cm. (Author).
Author: V. N. Shishov Publisher: ISBN: Category : Corrosion Languages : en Pages : 20
Book Description
In the search for more optimal core materials for a water cooled reactor at extended burnup, much attention is paid to alloys of the Zr-Nb and Zr-Nb-Fe-Sn systems. E110 and E635 alloys are two such. In the current VVER fuel cycle, the E110 alloy is used as fuel cladding and in SG components. The E635 alloy is under development as a fuel cladding and for fuel assembly structural elements for water cooled reactors of the VVER and RBMK types. E110, while having a unique corrosion resistance in pressurized water reactors, is subject to noticeable disadvantages in terms of corrosion resistance under conditions of boiling and higher coolant oxygen contents as well as in deformation stability under stresses and irradiation. Currently, the E635 alloy has passed the most important steps of qualification and is being introduced into cores as a material for guide thimbles, central tubes, and stiff frame angles in VVER-1000 FAA and FA-2. Properties of alloys are governed by their compositions and microstructure and even small changes in composition (Nb, Fe, Sn) and processing (heating in the ? or the ?+? regions) lead to substantial changes in properties as a result of changes in second phase precipitates and matrix composition. ATEM was used to study structure--phase states of a series of alloys Zr-(0.6-1.2) Nb-(0-0.6) Fe-(0-1.5) Sn (% weight), to determine the microstructural characteristics of recrystallized cladding tubes and the temperature stability regions of ?-Nb, ?-Zr, Zr(Nb,Fe)2, and (Zr,Nb)2Fe second phase precipitates. An increase in the relative content of iron R=Fe/(Fe+Nb) results in a larger volume fraction of (Zr,Nb)2 Fe precipitates. ?-Nb and Zr(Nb,Fe)2 particles are completely dissolved at ?750°C, the (Zr,Nb)2Fe phase at ?800°C. Autoclave corrosion tests revealed that the corrosion resistance of the materials depends on alloy composition. The content of tin lowered down to 0.8 % reduces weight gains in water, water containing Li, and particularly in steam. The content of Nb reduced to 0.6 % results in lower weight gains in water and steam and higher weight gains in Li containing water. The optimal content of iron in Zr-Nb-Fe-Sn alloys for corrosion resistance depends on the R ratio and makes up 0.2-0.4 %. Tests of samples produced from tubes of the above alloys and irradiated in BOR-60 at 315-345°C show that alloying Zr-Nb alloys with iron and tin improves their resistance to irradiation growth and creep. Sn and a higher Fe content in solid solution effected by transfer of Fe from the Laves phase precipitates to the matrix under irradiation strengthens the alloys. The influence of irradiation on phase compositions was established using irradiated samples (gas filled and unstressed) of cladding tubes: ?-Nb (85-90 % Nb) precipitates become depleted in niobium (or enriched in zirconium) to 50-60 % Nb and finely dispersed irradiation induced second particles (IIPs) enriched in niobium are formed. The Laves phase becomes depleted in iron and alters its crystal structure from hcp to bcc of the ?-Nb type. The fcc (Zr,Nb)2Fe precipitates retain on the whole their composition and structure, but the peripheries of particles reveal structural features, possibly related to niobium redistribution. No amorphization of any of the precipitates was identified. Alloy composition and applied stress under irradiation influence density and distribution of dislocation loops and IIP precipitates. Proceeding from results of out-of-pile and from post-irradiation examinations of the structure and properties of E110 and E635 type cladding tubes, compositions of alloys having improved corrosion and irradiation resistances are proposed. E110 type (Zr-1Nb-0.1Fe-0.1O) alloy features enhanced strength characteristics as a result of iron transfer from Laves phase precipitates to the matrix under irradiation, lower irradiation induced growth strain, and irradiation-thermal creep. An E635 type alloy (tin and niobium content lowered down to
Author: Marc Tupin Publisher: ISBN: Category : Corrosion Languages : en Pages : 41
Book Description
The irradiation damage in the fuel cladding material is mainly caused by the neutron flux resulting from the fission reactions occurring in the fuel. From an experimental point of view, the neutrons have the disadvantage to activate materials by neutron capture rendering them difficult to handle. To avoid these constraints inherent in the handling of radioactive material, the radiation effects on the corrosion resistance of zirconium alloys can be studied by irradiating the materials with ions. A new experimental approach using ion irradiation was performed in the Microscopy and Irradiation Damage Studies Laboratory of the CEA in Saclay, with the aim to study more specifically the influence of the irradiation damages in the oxide on the corrosion rate of the zirconium alloys. This study was, moreover, focused on a particular distribution of defects in the oxide layer, basically, localised close to the metal/oxide interface. From the results of the irradiation of the metal/oxide interface, it was clearly shown that, whatever the incident ion, the irradiation of the internal interface results in a significant increase of the oxygen diffusion flux ratios between the most irradiated Zircaloy-4 and the unirradiated one, whereas that of the oxide formed on M5TM induces a big decrease of the oxygen diffusion flux in the film. These effects are less marked with helium ions compared to protons (M5TM is a trademark of AREVA NP registered in the United States and in other countries). Finally, the oxide irradiation impact on the oxygen diffusion through the layer could explain the corrosion acceleration factor observed on Zy4 during the first cycles of irradiation, but cannot alone explain observed corrosion accelerations under high burn-up conditions. The discussion on the oxide irradiation effects puts forward the probable role of the residual charge left by ion implantation.
Author: Holger Wiese Publisher: ISBN: Category : High burnup Languages : en Pages : 34
Book Description
The objective of this paper is to summarize the results of the latest observations performed at Paul Scherrer Institut on irradiated fuel claddings, to characterize their corrosion and hydrogen-uptake behavior. Two categories of studies have been performed. (1) A series of destructive tests were achieved on the fuel rods irradiated in a boiling-water reactor (BWR), including hydrogen concentration by hot-gas extraction. These results provided the hydrogen content of the cladding at different stages of irradiation, at different elevations along the rod. (2) Another series of examinations using a correlative microscopy method, i.e., using different techniques, including transmission electron microscopy (TEM), electron probe microanalysis (EPMA), and secondary ion mass spectrometry (SIMS), on the same material and in the same region of the metal-oxide interface have provided useful data regarding the oxide layer combining the signals from oxides and from hydrides. Furthermore, the effect of the type of alloying element has been examined for in-reactor oxidation. These studies are subsequently combined with the findings from out-of-pile studies, using techniques, such as neutron radiography, to confirm the in-reactor observations. Results have shown that: (i) the hydrogen pickup fraction varies at different conditions and could even decrease as the oxide thickness increases; (ii) the distribution of hydrogen in the cladding is usually inhomogeneous; (iii) the most determining parameter for hydrogen uptake seems to be the microstructure of the oxide, and the nature of the alloying element will influence to a certain extent this parameter; (iv) furthermore, the stress in the oxide layer can modify the crack distribution in the latter, cracks will in turn shorten the route for the hydrogen to access the metal. These results will be discussed as a contribution to the available knowledge about hydrogen uptake and will provide a global support for the models of the uptake phenomenon.
Author: Zefeng Yu Publisher: ISBN: Category : Languages : en Pages : 276
Book Description
Zirconium-based alloys have been used in nuclear fission reactors, because of their low thermal neutron cross-section, good mechanical strength, and adequate corrosion resistance. In pressurized water reactors, one of the major reasons that Zr-Nb alloys have widely replaced Zircaloys-4 is the absence of the accelerated oxide growth at high burnup. Although such great advantages have led to the development of many advanced commercial Zr-Nb alloys, the reasons behind the enhanced in-reactor corrosion resistance are still unclear. The distribution and concentration of alloying elements in the substrate have been suspected bo play a major role in controlling in-reactor corrosion kinetics. The major hypothesis being tested in this thesis study is that the enhanced corrosion resistance of irradiated Zr-Nb alloys is a result of irradiation-induced reduction of Nb concentration in solid solution, due to the nucleation and growth of Nb-rich irradiation-induced platelets (IIPs)/nanoclusters. To validate our hypothesis, a systematic study has been performed to understand the irradiation induced microstructure and microchemistry evolution and the subsequent effect on the corrosion kinetics of Zr-xNb (x=0.2, 0,4, 0,5, 1.0) model alloys. The microstructure and microchemistry of as-received materials were characterized under STEM/EDS/APT. Then, 2 MeV proton irradiation has been performed on these model alloys at 350 °C up to 1 dpa. (S)TEM/EDS has been used to study the size and density evolution of the native precipitate and IIPs as a function of irradiation dose. The major use of APT is to quantify the Nb concentration in the solid solution as a function of irradiation dose in order to support our hypothesis. The IIPs crystal structure and growth mechanism have been particularly inverstigated using HRSTEM and 4D-STEM. After the characterization, the irradiated materials were corroded in autoclave to study if the proton irradiation leads to subsequent lower corrosion rate. Lastly, the same characterization techniques and methods have been used to study neutron irradiated commercial alloys, M5®, ZIRLO® and X2®, in an effort to compare the results with proton irradiation. The possible IIPs nucleation and growth mechanism and the effects of irradiation-induced Nb redistribution on the corrosion kinetics are the major focuses of the discussion section.
Author: VF. Urbanic
Author: VF. Urbanic
Author: BD. Warr
Author: VF. Urbanic
Author: Frederic R. Shober
Author: V. N. Shishov
Author: V. F. Urbanic
Author: Marc Tupin
Author: Holger Wiese
Author: Zefeng Yu
Author: Glenn Herbert Jenks