Advancements in transmission electron microscopy allow us to draw correlations between evolving matrix chemistry environments and the resulting dislocation structures that form. Such phenomena are ...essential in predicting the lifetime of neutron reactor components, but are not well understood at the fundamental level. We investigate the effect of nano-scale matrix chemical evolution in Zircaloy-2 on dislocation formation after emulating commercial reactor irradiation conditions on a proton beamline. Similarity in the dislocation type, morphology, density and evolution between the different irradiation types establishes proton irradiation in this regard. For the first time, we observe chemical segregation of Fe, Ni and Cr to a-loop positions in basal traces and the segregation of Sn in alternate rows, anticorrelated to the positions of the light transition elements. The resulting layered structure with a periodicity of ∼50 nm creates an even greater anisotropy than that usually associated with HCP materials. Concurrent analysis of chemical effects and dislocation spatial relationships provides evidence that may explain the delayed onset of c-loop nucleation and accelerated dimensional instability regimes in its dependence on the alignment of a-loops parallel to the trace of the basal plane. This demonstrates the applicability of chemical-structural correlations towards key research questions regarding deformation behaviour.
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Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak ...profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4–15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys.
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Detailed analysis was carried out on proton and a neutron irradiated Nb-containing Zr-alloy to study the evolution of dislocation loop size and densities as well as the formation and evolution of ...irradiation-induced precipitation/clustering. The results obtained here have been contrasted against previously published work on a Nb-free Zr-alloy 1, 2 to investigate the mechanistic reason for the improved resistance to irradiation-induced growth of Nb-containing Zr alloys. The combined use of bright field scanning transmission electron microscopy, ultra-high-resolution energy dispersive spectroscopy and atom probe tomography analysis provides evidence of evenly distributed radiation-induced Nb clusters that have formed during the early stage of proton irradiation and Fe-rich nano-rods near Fe-containing second phase particles. The former seems to have a profound effect on loop and subsequent loop formation, keeping loop size small but number density high while loops seem to initially form at similar dose levels compared to a Nb-free alloy but loop line density does not increase during further irradiation. It is hypothesized that the formation of the Nb nano-precipitates/clusters significantly hinders mobility and growth of loops, resulting in a small size, high number density and limited ability of loops to arrange along basal traces compared to Nb-free Zr-alloys. It is suggested that it is the limited loop arrangement that slows down loop formation and the root cause for the high resistance of Nb-containing Zr-alloys to irradiation-induced growth.
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The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of ...long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). It was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.
Transmission electron microscopy (TEM) studies provide mechanistic understanding of nanoscale processes, whereas advanced synchrotron XRD (SXRD) enables precise measurements on volumes that are more ...representative of bulk materials. Therefore, the combined strengths of these techniques can provide new insight into irradiation-induced mechanistic processes. In the present study, their application to Zircaloy-2, proton-irradiated to 2.3, 4.7, and 7.0 dpa at 2 MeV and 350 °C and neutron-irradiated to 9.5 and 13.1 × 1025 n m−2 are exemplified. The application of correlative spectral imaging and structural TEM investigations to the phase transformation of Zr(Fe,Nb)2 precipitates in Low-Sn ZIRLO™, neutron-irradiated to 8.9–9 × 1025 n m−2, demonstrates the possibility of a Cr core nucleation site. Anomalous broadening is observed in SXRD profiles, which is believed to be caused by defect clusters and precursors to dislocation loop nucleation. The challenges to quantitative analysis of dislocations by SXRD are highlighted with reference to the segregation of Fe and Ni to basal planes and dislocation cores, observed by spectral imaging in the TEM.
•A novel gamma transmission densitometer is presented for post-irradiation examination of nuclear fuel.•The densitometer performance was characterized by a dedicated test campaign in a hot cell ...laboratory.•For a set of calibration standards, the local density was measured with a discrepancy of few percent.•A spatial resolution of a few hundred microns was obtained and quantified using an edge spread test.•The density profile of an irradiated accident-tolerant fuel sample was measured to demonstrate the measurement concept.
Collimated Gamma Transmission Micro-Densitometry (GTMD) is a novel technique proposed to investigate local density variations of nuclear fuel in PIE, with a high spatial resolution. In this work, the first experimental tests of a gamma micro-densitometer are presented and the performance is characterized. The experimental procedures are described, including the aligning process and the calibration methodology. The results demonstrated that for the calibration samples with a thickness above 5 mm, a local density was obtained with a maximum discrepancy of about 2% and a spatial resolution of about 280 µm. The setup was used for the first test on an irradiated ADOPTTM fuel pellet slice. From the measurement, an average bulk density of about 9.58 g/cm3 was calculated and local density features were observed, possibly related to rim effects or the presence of local cracks. The information acquired also presented valuable information for possible improvements in the setup’s performance.
This paper presents simulations of the xM3 power ramp with the fuel performance code ALCYONE performed during an international simulation exercise organized within the Organisation for Economic ...Co-operation and Development/Nuclear Energy Agency Power to Melt and Maneuverability project. The xM3 test involved a large-grain UO
2
fuel from Mitsubishi Heavy Industries cladded with Zirlo and pre-irradiated in a Spanish pressurized water reactor up to an average burnup of 27 GWd/tU
−1
. It was then submitted to a staircase ramp protocol in the R2 reactor at Studsvik (Sweden) with 10 successive steps of 5 kW·m
−1
up to a ramp terminal level of 70 kW·m
−1
. The fuel rodlet did not fail, and detailed post irradiation examinations performed during the Studsvik Cladding Integrity Project II evidenced recrystallization of the pellet center around a central hole, interpreted as signs of fuel melting.
In this paper, simulations with ALCYONE of the xM3 power ramp, including an advanced model for fuel melting based on thermodynamic equilibrium calculations, are detailed. The model relies on the determination of the liquid fuel fraction evolution with temperature that is used to obtain a continuous description of the material properties during phase change. In consequence of the incorporation of rare earths and actinides in the bulk of the fuel, distinct solidus and liquidus temperatures are estimated. It is shown that the observed central hole and recrystallized central part of the pellet could be the consequence of totally melted fuel (liquidus is reached), partially melted fuel (solidus is reached), or pore migration only.
A single hydride platelet and the matrix material next to it in a Zircaloy-2 cladding have been targeted for hardness, H, and Young’s modulus, E, measurement using nanoindentation. The results were ...compared with those obtained in the matrix material far away from the hydride.
The results show that hardness and Young’s modulus in the hydride are higher than those of the matrix adjacent to the hydride, which are the same as those of the matrix far away from the hydride.
Zirconium alloys used as cladding materials in nuclear reactors can exhibit accelerated irradiation induced growth, often termed linear growth, after sustained neutron irradiation. This phenomenon ...has been linked to the formation of -component dislocation loops and to the concentration of interstitial solute atoms. It is well documented for the Zircaloys that Fe dissolves from second phase particles (SPPs) during irradiation thus increasing the interstitial solute concentration in the matrix. However, no progress has yet been made into understanding whether a similar process occurs for the newer ZIRLO™ alloys. We aim to overcome this shortcoming here by studying compositional changes in second phase particles in Low Tin ZIRLO™ after neutron and proton irradiation using energy dispersive X-ray (EDX) spectroscopy. Material irradiated to 18dpa (displacements per atom) using neutrons and to 2.3 and 7dpa by protons was investigated. The results show that Fe is lost from Zr–Nb–Fe-SPPs during both neutron and proton irradiation. Prior to irradiation, Fe was detected at the interface of β-Nb-SPPs. This Fe enrichment is also dispersed during irradiation. Qualitatively, excellent agreement was found regarding the elemental redistribution processes observed after proton and neutron irradiation.
Proton-and neutron-irradiated Zircaloy-2 are compared in terms of the nano-scale chemical evolution within second phase particles (SPPs) Zr(Fe,Cr)2 and Zr2(Fe,Ni). This is accomplished through ...ultra-high spatial resolution scanning transmission electron microscopy and the use of energy-dispersive X-ray spectroscopic methods. Fe-depletion is observed from both SPP types after irradiation with both irradiative species, but is heterogeneous in the case of Zr(Fe,Cr)2, predominantly from the edge region, and homogeneously in the case of Zr2(Fe,Ni). Further, there is evidence of a delay in the dissolution of the Zr2(Fe,Ni) SPP with respect to the Zr(Fe,Cr)2. As such, SPP dissolution results in matrix supersaturation with solute under both irradiative species and proton irradiation is considered well suited to emulate the effects of neutron irradiation in this context. The mechanisms of solute redistribution processes from SPPs and the consequences for irradiation-induced growth phenomena are discussed.
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•Protons emulate the effects of neutron irradiation in the evolution of chemistry and morphology of second phase particles.•Detailed energy-dispersive X-ray spectroscopy reveals heterogeneity in Zr-Fe-Cr SPPs both before and after irradiation.•Zr-Fe-Ni SPPs are delayed in irradiation-induced dissolution due to their better self-solubility with respect to Zr-Fe-Cr.
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