Hafnium corrosion behavior in high-temperature steam

Corrosion properties for three kinds of hafnium have been examined in 10.5 MPa steam at 773 K. Nuclear grade hafnium formed a shiny black film with a few white nodules, which were found to be monoclinic HfO₂. Hydrogen pick-up fraction during the corrosion amounted to ca. 40% to produce hydrogen dissolved in the hafnium matrix with evolving the remaining 60% hydrogen as H₂ gas. Sponge and crystal bar hafnium were also examined and they showed superior and inferior corrosion resistance, respectively, to nuclear grade hafnium. The corrosion resistance for hafnium increased with increasing the iron content involved in hafnium for these three kinds of hafnium and, correspondingly related to the density of iron-containing second phase particles, which were characterized to be face centered cubic Hf₂Fe.The corrosion mechanism, which was previously proposed for Zircaloy nodular corrosion, was adopted with making minor alterations, to explain the hafnium corrosion properties.     

Influence of prior dislocation structure on anisotropy of thermal creep of cold-worked Zr-2.5Nb tubes

Normally, creep anisotropy of hcp metals is thought to be controlled by the crystallographic texture. Here, we show that the creep anisotropy of cold-worked Zr-2.5Nb tubes is also very dependent on the anisotropic dislocation structures introduced by cold-work. The contribution of each slip system to the creep deformation of an individual grain orientation depends upon the activity of that slip system during prior cold-work. This conclusion is reached by comparing the self-consistant visco-plastic polycrystalline models with thermal creep tests performed on internally pressurized thin-wall capsules with different textures under a transverse stress of 300 MPa at 350 °C, where dislocation creep is the dominant operating mechanism. The non-uniform dislocation distributions prior to creep were derived by simulating the cold-work process of Zr-2.5Nb tubes from an Elasto-Plastic Self-Consistent (EPSC) model.     

Contribution of uranium diffusion on creep behavior of uranium dicarbide

Compressive creep tests of uranium dicarbide (UC₂) have been conducted. The general equation best describing the creep rate over the temperature range 1200–1400°C and over the stress range 2000–15000 psi is represented by the sum of two exponential terms ge =A(σ/E)^0.9 exp(−39.6 ± 1.0/RT) + B(σ/E)^4.5 exp(−120.6 ± 1.7/RT), where pre-exponential factors are A(σ/E)^0.9 = 12.3/h at low stress region (3000 psi) and B(σ/E)^4.5 = 3.17 × 10¹³/h at high stress region (9000 psi), and the activation energy is given in kcal/mol. Each term of this experimental equation indicates that important processes occurring during the steady state creep are grain-boundary diffusion of the Coble model at low stress region and the Weertman dislocation climb model at high stress region. Both mechanisms are related to migration of uranium vacancies.     

Effects of helium on the high-temperature creep and fracture properties of Ni-6% W

High-temperature constant true stress creep tests have been perfomed on samples of Ni-6% W which were injected with α-particles to calculated helium concentrations of 0.4 ppm and 4.0 ppm. The apparent activation energies for creep were determined to be between 260 kJ/mol at 20.7 MPa and 364 kJ/mol at 48.2 MPa and were found to be greater than 247 kj/ mol, the activation energy for creep in similar samples which did not contain helium. These high activation energies for the injected material are believed to be fictitious in the sense that they are caused by the effects of helium on the shape of the creep curve. The helium bubbles present at grain boundaries promote premature fracture early in the primary creep stage by lowering the effective fracture-surface energy and thereby prevent the observation of steady-state creep in the injected material.     

Influence of the structure and properties of uranium on its behavior during irradiation

Data are presented on the variations in dimensions and form of uranium specimens during irradiation. It is shown that by regulating the composition of the uranium and treatment conditions (degree of deformation in the-region and heat-treatment conditions), in consequence of variation in grain size and texture, it is possible to vary within wide limits the magnitude of surface distortion due to irradiation and the value of G_i.A study has been made of the dependence of the variation in grain size of quenched uranium, as well as hardness, tensile strength, and yield strength, on the iron, silicon, and aluminum content of uranium. The cooling rate and content of these impurities influence the critical point of the transformation on quenching; for example, for a cooling rate of 400C/ sec and a silicon content of 0.05%, the critical point of the transformation drops to 530C.Experimental results show a creep acceleration during irradiation (nv = 6·10¹² neutr/cm²·sec) of 50–100 times, i.e., by 1.5–2 orders for textured uranium and uranium with disoriented structure. The rate of creep of uranium with a disoriented structure is connected to the burnup rate.The results are given of tensile tests made on uranium directly in the reactor. Even after remaining a short time in the neutron field (up to 1 hour), the percentage elongation is diminished somewhat and the tensile strength is increased.The following assisted in the experimental work: A. G. Lanin, V. M. Teplinskaia, V. K. Zakharova, L. N. Protsenko, V. N. Golovanova and K. A. Borisov.     

Hydride behavior in Zircaloy cladding tube during high-temperature transients

In order to study the hydride behavior in high burnup fuel cladding during temperature transients expected in anticipated operational occurrences and accidents, unirradiated hydrided Zircaloy-4 cladding tubes were rapidly heated to temperatures ranging from 673 to 1173 K and annealed for holding time ranging from 0 to 3600 s. Hydrides were localized in the peripheral region of the cladding tubes prior to the annealing, as observed in high burnup fuel cladding. The localized hydride layer (hydride rim) was annealed out, and the radial hydride distribution became uniform after the annealing at 873 K for 600 s, 973 K for 60 s, or 1173 K for 0 s. The annealing out of the hydride rim is caused by the phase transformation from the α + δ phase to the α + β or β phase in the hydride rim and the subsequent drastic increase in the solid solubility and diffusion of hydrogen in Zircaloy. On the other hand, the radial distribution and morphology of hydrides did not change at lower temperatures: Thus, the hydride remains almost intact below the phase transformation temperature for the short time ranges.     

Simulation of molten metal freezing behavior on to a structure

In the severe accident analysis of liquid metal reactors (LMRs), understanding the freezing behavior of molten metal onto the core structure during the core disruptive accidents (CDAs) is of importance for the design of next-generation reactor. CDA can occur only under hypothetical conditions where a serious power-to-cooling mismatch is postulated. Material distribution and relocation of molten metal are the key study areas during CDA. In order to model the freezing behavior of molten metal of the postulated disrupted core in a CDA of an LMR and provide data for the verification of the safety analysis code, SIMMER-III, a series of fundamental experiments was performed to simulate the freezing behavior of molten metal during penetrating onto a metal structure. The numerical simulation was performed by SIMMER-III with a mixed freezing model, which represents both bulk freezing and crust formation. The comparison between SIMMER-III simulation and its corresponding experiment indicates that SIMMER-III can reproduce the freezing behavior observed on different structure materials and under various cooling conditions. SIMMER-III also shows encouraging agreement with experimental results of melt penetration on structures and particle formation.     

Simulation of high-temperature swelling of uranium dioxide and deformation behavior of a fuel element

Models and computer codes, developed based on them, for simulating the swelling of uranium dioxide (BARS) and the stress-deformation state of a fuel element (SDS) under high-temperature irradiation are presented. It is shown that when developing a design for high-temperature fuel elements and validating their serviceability the quantitative indicator required for the swelling of uranium dioxide in the range ≥1400°C is the change in the external dimensions of the fuel caused by constant formation and growth of bubbles containing gaseous fission products during irradiation. The results of computational investigations using the models indicated are examined. These results eliminate the inconsistency of the data on the effect of the main operating parameters — the temperature and burnup — on the radiation characteristics and service life behavior of a fuel element. It is shown that the central channel in the fuel kernel and strengthening of the cladding improve the dimensional stability fuel elements. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 172–179, September, 2007.     

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective e?ect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.     

Martensitic zirconium alloys: Influence of chemical composition on creep characteristics

Results of an extensive investigation of creep in martensitic zirconium alloys are summarized with the aim to show the influence of chemical composition on the main creep characteristics — the steady state creep rate and the time and strain to fracture. The activation energy of creep and the parameter of stress sensitivity of steady state creep rate are determined and possible creep mechanisms as well as creep strengthening mechanisms are discussed. The time to fracture $\text{t}{}_{\mathrm{f}}$, is related to the steady state creep rate ge_s through the Monkman-Grant relation as modified by Dobe? and Milic?ka. The creep fracture shows features different from those of “classical” intergranular cavitation creep fracture. Most probably the creep fracture is controlled by the same deformation mechanism as the creep.     

The various creep models for irradiation behavior of nuclear graphite

Graphite is a widely used material in nuclear reactors, especially in high temperature gascooled reactors (HTRs), in which it plays three main roles: moderator, reflector and structure material. Irradiation-induced creep has a significant impact on the behavior of nuclear graphite as graphite is used in high temperature and neutron irradiation environments. Thus the creep coefficient becomes a key factor in stress analysis and lifetime prediction of nuclear graphite. Numerous creep models have been established, including the visco-elastic model, UK model, and Kennedy model. A Fortran code based on user subroutines of MSC.MARC was developed in INET in order to perform three-dimensional finite element analysis of irradiation behavior of the graphite components for HTRs in 2008, and the creep model used is for the visco-elastic model only. Recently the code has been updated and can be applied to two other models—the UK model and the Kennedy model. In the present study, all three models were used for calculations in the temperature range of 280–450 °C and the results are contrasted. The associated constitutive law for the simulation of irradiated graphite covering properties, dimensional changes, and creep is also briefly reviewed in this paper. It is shown that the trends of stresses and life prediction of the three models are the same, but in most cases the Kennedy model gives the most conservative results while the UK model gives the least conservative results. Additionally, the influence of the creep strain ratio is limited, while the absence of primary creep strain leads to a great increase of failure probability.     

Tertiary creep behavior of annealed 214 Cr-1 Mo steel

Under various creep conditions for annealed $\text{2}\text{1}\text{4}$ Cr-1 Mo steel, nonclassical creep curves that contain two steady-state stages were observed. The transition from the first to second steady-state stage involves a quasi-tertiary (increasing creep rate) stage, thus complicating the definition of tertiary creep. Tertiary creep is important because it is often associated with the formation of gross structural instability (i.e., the formation of cracks, voids, or a neck). The present studies indicated a consistent correlation between the onset of tertiary creep and rupture life was obtained when the end of the second steady-state stage was used as the onset of tertiary creep for the nonclassical curves. The creep strains to the end of the second steady-state stage were similar to those to the end of the secondary stage of the classical curves. These results along with previous work indicate that the creep rate during the second steady-state stage of the nonclassical curves is controlled by the same process that controls creep during the secondary stage of a classical curve.     

The effects of temperature and dose-rate variations on the creep of austenitic stainless steels in the dounreay fast reactor

Loaded helices of cold-worked, austenitic stainless steels (En58B, E, J, FV548 and type 316) have been irradiated in DFR at damage rates ranging over an order of magnitude at temperatures between 513 and 633 K. Creep rates varied between steels but a common pattern emerged where strain per unit dose appeared to increase with decreasing dose-rate. Comparison with experiments at 773 K on three of the steels reveals creep rate increases by factors ranging from 1.2 to 5. The results are discussed in the context of other fast reactor data and possible explanations of the findings are considered.     

Numerical investigations of thermal mixing performance of a hot gas mixing structure in high-temperature gas-cooled reactor

A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in the hot gas chamber and the hot gas duct of the HTR were obtained based on the commercial computational fluid dynamics(CFD) program. The numerical simulation results showed that the helium flow with different temperatures in the hot gas mixing chamber and the hot gas duct mixed intensively, and the mixing rate of the temperature in the outlet of the hot gas duct reached 98 %. The results indicated many large-scale swirling flow structures and strong turbulence in the hot gas mixing chamber and the entrance of the hot gas duct, which were responsible for the excellent thermal mixing of the hot gas chamber and the hot gas duct. The calculated results showed that the temperature mixing rate of the hot gas chamber decreased only marginally with increasing Reynolds number.     

Influence of loading rate on the fracture toughness versus temperature curve

This paper presents recent experimental investigations on the influence of loading rate on the fracture toughness K_Ic for different structural steels. The loading rate in terms of increasing in stress intensity factor was changed from quasistatic up to dynamic conditions (

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). The results show significant differences in the amount of temperature shift between K_Ic -T curve obtained after static and dynamic loading for the materials investigated. Based on microscopic fracture criteria for cleavage fracture correlations were made between fracture toughness and yielding behaviour depending on temperature and strain rate. The experimental results were also compared with the predictions given by different models. The most promising results were achieved by a correlation between transition temperature shift in K_Ic caused by dynamic loading and strain rate sensitivity of steels.     

Influence of cyclic temperature transients on the creep behaviour of heat exchanger tubes

Due to thermal fluxes hot streaks exist in the coolant media of heat exchanger components. They cause alternating cyclically secondary stresses in the component walls which superpose on the primary stresses due to internal pressure or bending. Experimentally it was shown that hot streaks at high temperatures influence the creep behaviour very strongly. Dependent on the ratio of primary and secondary stresses the creep rate of the components is higher than the creep behaviour at the highest cycle temperature under primary stresses only.