Lanthanides migration in U-Zr based nuclear fuels

Atomistic modeling using the BFS method for alloys is performed to study the formation of lanthanide-rich precipitates in U-Zr fuel and the segregation patterns of all constituents to the surface. Surface energies for all elements were computed and, together with the underlying concepts of the BFS method, the migration of lanthanides to the surface region in U-Zr fuels is explained.     

Gamma-Ray Production Cross Sections for Interactions of 14.8 MeV Neutrons with O,Na, Al,Cl, Cr,Fe, Ni,Cu and Pb

Production cross sections for 85 discrete γ-rays at 125° were measured with a Ge(Li) detector for interactions of 14.8 MeV neutrons with natural samples of O, Na, Al, Cl, Cr, Fe, Ni, Cu and Pb. The obtained cross sections were compared with the results of previous works. For O, Na, Al, Cr and Ni, the present results agree with the previous data measured with monoenergetic neutron sources; for Cl, Fe, Cu and Pb, the present results are larger than the previous data. In comparison between the present results shown by histograms of γ-ray energy and unfolded data, a considerable discrepancy is found from some of the previous data for Fe.     

Density-functional study of Zr-based actinide alloys

Density-functional formalism is applied to study the phase equilibria in the U-Zr system. The obtained ground-state properties of the γ (bcc) and δ (C32) phases are in good agreement with experimental data. The decomposition curve for the γ-based U-Zr solutions is calculated. We argue that stabilization of the δ-UZr₂ phase relative to the α-Zr (hcp) structure is due to an increase of the Zr d-band occupancy that occurs when U is alloyed with Zr.     

Effect of impurity and alloying elements on Zr grain boundary strength from first-principles computations

Effects of twenty impurity and alloy elements on the strength of a Zr(0 0 0 1)/Zr(0 0 0 1) ∑7 twist grain boundary were studied using a first-principles density functional approach. A ranking in the order of most weakening to most strengthening was: Cs, I, He, Te, Sb, Li, O, Sn, Cd, H, Si, C, N, B, U, Ni, Hf, Nb, Cr, and Fe. Segregation energies for these elements to the grain boundary and the Zr(0 0 0 1) surface were also calculated. Calculations showed that the weakening grain boundary elements He, I, and Cs have a strong driving force for segregation to the grain boundary from bulk Zr. Zircaloy cladding failures (pellet-clad interactions) in commercial fuel systems and separate effects test results provide context for these computational results.     

Zr（Fe,Cr）2金属间化合物的氧化

用非自耗电弧炉熔炼了比值(重量比值)不同的Zr(Fe,Cr)_2,并在773K和973K的空气中氧化。经X射线衍射和电子衍射分析表明:当Fe/Cr≤4.5时,Zr(Fe,Cr)_2,是MgZn_2型(六方)的Laves相,它的晶格常数随Fe/Cr比增加而收缩。Zr(Fe,Cr)_2氧化后生成的稳定氧化物是单斜ZrO_2和六方(Fe,Cr)_2O_3。在形成稳定氧化物之前,还会出现亚稳定的立方ZrO_2。根据本实验结果讨论了Zr-4合金中Zr(Fe,Cr)_2第二相对腐蚀性能的影响。     

Performance of FCCI barrier foils for U-Zr-X metallic fuel

Diffusion couple tests of U-Zr or U-Zr-Ce alloys vs. ferritic martensitic steels such as HT9 or T91 were carried out in order to evaluate the performance of the diffusion barrier candidates. Elemental metal foils of Zr, Nb, Ti, Mo, Ta, V and Cr were very effective in inhibiting interdiffusion between these fuels and steels. Eutectic melting between the fuels and steels was not observed in any of the diffusion couples using these diffusion barrier foils at annealing temperatures up to 800 °C. Among the metallic foils evaluated in this study, V and Cr exhibited the most promising performances as a diffusion barrier material for eliminating the fuel cladding chemical interaction problem. However, Zr, Nb and Ti showed an active interaction with the fuel mainly due to the large U solubility.     

Effect of Zirconium Addition to Austenitic Stainless Steels on Suppression of Radiation Induced Chromium Segregation at Grain Boundaries under Ion Irradiation

The effect of Zr addition to austenitic stainless steels on the suppression of radiation induced Cr segregation at grain boundaries under 400 keV He⁺ irradiation was studied. Type 316L stainless steel and steels with addition of 0.07, 0.21 or 0.41 mass% Zr were kept at 1,423K for 30 min, and then they were quenched into the water. Irradiation was done at 773K with the dose rate of 2.4×10^?4dpa/s. The total dose was 0.85 or 3.4dpa. After irradiation, profiles of Cr concentration across the grain boundaries were measured using an analytical electron microscope with 1 nm beam diameter. Concentration of Cr at the grain boundary is decreased by radiation induced segregation. However, it increased with the addition of Zr, and the Cr segregation is almost completely suppressed when Zr is added more than 0.21 mass%.

The effect of Zr addition on suppression of Cr segregation was analyzed focussing on the interaction between dissolved Zr atoms and point defects. The effect is based on vacancy trapping by the Zr atom, and the extent to which it suppresses Cr segregation can be empirically evaluated using a radiation induced segregation model by changing the effective vacancy migration energy.     

In situ oxidation of zirconium binary alloys by environmental SEM and analysis by AFM, FIB, and TEM

Binary Zr-alloys containing 1%Fe and 1% Ni (large precipitates) and 1% Cr and 0.6% Nb (small precipitates), as well as a pure Zr sample were exposed in situ at 130 Pa water vapour pressure at 415 °C in an environmental SEM. The surface topography and composition of each sample was characterised before in situ experiments, during and after oxidation. After oxidation the surface was characterised by SEM and EDS, AFM and TEM combined with EDS. Focused ion beam was used to prepare cross sections of the metal-oxide interface and for the preparation of TEM thin foils.The oxidation behaviour of precipitates for these alloying elements can be characterised into two large families, those which show a rapid oxidation and those which induce a delayed oxidation in comparison with the Zr-matrix.At 415 °C after 1 h of oxidation for Zr1%Fe and Zr1%Ni, the formation of protrusions could be detected at the surface, being related to underlying SPP in the oxide. On Zr1%Cr and Zr0.6%Nb unoxidised SPPs were observed in the oxide, close to the metal-oxide interface. These SPPs were, however, oxidised close to the outer surface of the oxide. The surface roughness was increased for all materials after in situ oxidation, however, only for Zr1%Fe and Zr1%Ni protrusions appeared on the surface during oxidation. It was subsequently demonstrated that these latter correspond to the position of SPPs. For Zr1%Fe the surface roughness increased more than in the other materials and on these protrusions small iron oxide crystals have been observed at the surface. These observations confirm that Fe has a different behaviour compared to the other SPP forming elements, and it diffuses out to the free surface of the material.These alloying elements being the constituents of the commercial alloys (Fe and Cr for Zircaloy-4; Fe, Cr and Ni for Zircaloy-2 and Nb for all Nb-containing alloys), this study allows to separate their individual influence and can allow a subsequent comparison to the behaviour of those more complex alloys.     

Metallurgical characterizations of Fe–Cr–Ni–Zr base alloys developed for geological disposal of radioactive hulls

Alloy melting route is currently being considered for radioactive hulls immobilization. Towards this, wide range of alloys, belonging to Zirconium–Iron binary and Zirconium–Stainless steel pseudo-binary systems have been prepared through vacuum arc melting route. Detail microstructural characterization and quantitative phase analyses of these alloys along with interaction study between Zirconium and Stainless steel coupons at elevated temperatures identify Zr(Fe,Cr)₂, Zr(Fe,Cr), Zr₂(Fe,Cr), Zr₃(Fe,Ni), Zr₃(Fe,Cr), Zr₃(Fe,Cr,Ni), β-Zr and α-Zr as the most commonly occurring phases within the system for Zirconium rich bulk compositions. Nano-indentation studies found Zr(Fe,Cr)₂ and Zr(Fe,Cr) as extremely hard, Zr₃(Fe,Ni) as moderately ductile and β-Zr, Zr₂(Fe,Cr) as most ductile ones among the phases present. Steam oxidation studies of the alloys, based on weight gain/loss procedure and microstructural characterization of the mixed oxide layers, suggest that each of the alloys responded to the corrosive environment differently. Fe₂O₃, NiFe₂O₄, NiO, monoclinic ZrO₂ and tetragonal ZrO₂ are found to be most common constituents of the oxide layers developed on the alloys. Integrating the microstructural, mechanical and corrosion properties, ZrFeCrNi3 (Zr: 84.00, Fe: 11.20, Cr: 3.20, Ni: 1.60, in wt.%) is identified as the acceptable base alloy for disposal of radioactive hulls.     

U-10Mo/Zr单片式燃料元件堆内稳态热-力学性能研究

本文建立了U-10Mo/Zr单片式燃料元件的辐照性能模型以及热-力学本构关系,采用有限元方法进行非均匀辐照场中燃料元件稳态热-力学性能的数值模拟,获得并分析了U-10Mo/Zr单片式燃料元件温度、形变和应力的分布特点及变化规律。研究结果表明,燃料芯体厚度增量在芯体和包壳结合面附近达到最大,主要受到燃料辐照蠕变的影响;在较低燃耗条件下,燃料芯体高温辐照肿胀模拟结果与低温辐照肿胀试验结果相当;燃料芯体边角区域和包壳端面外侧区域存在应力集中。      

Stability of the body-centered cubic gamma phase in the uranium-zirconium-niobium system

A thermodynamic representation of the stability of the binary gamma phases has been developed for the U-Zr, U-Nb and Zr-Nb binary systems. This has permitted the calculation of the miscibility gaps in the ternary system. The three binary gaps merge to form a large three-phase region in the middle of the composition trangle. Because of the large coherency energies, the spinodal decomposition in the U-Nb and Zr-Nb gaps will be drastically modified, but a large region remains in the ternary system where the decomposition products have small coherency energies such that spinodal decomposition is probable. From these results, we have identified the four decomposition processes in Mulberry (U-16.6 at% Nb-5.6 at% Zr) as being discontinuous and spinodal decomposition within the miscibility gap, the diffusionless formation of α and the formation of (α + γ) from the supersaturated gamma phase.     

The effect of Cr concentration on radiation damage in Fe-Cr alloys

Displacement cascades in Fe-Cr alloys were studied using molecular dynamics computer simulations. We considered random Fe-5Cr and Fe-15Cr alloys, as well as Fe-10Cr alloys with and without Cr-rich precipitates. In the simulations two versions of a two-band embedded atom method potential were used, and the cascades were induced by recoils with energies up to 20 keV. We found that the average number of surviving Frenkel pairs and the fraction of vacancies and self-interstitials in clusters was approximately the same in pure Fe and random Fe-Cr alloys (regardless of Cr concentration). A noticeable effect of the presence of Cr in the Fe matrix was only observed in the enrichment of self-interstitials by Cr in Fe-5Cr. The calculated change in the short range order parameter showed that Fe-5Cr tends towards ordering (negative short range order parameter) and Fe-15Cr towards segregation (positive short range order parameter) of Cr atoms. In simulations with the Cr-rich precipitate, enhanced cascade splitting and segregation of self-interstitial defects created inside the precipitates towards the precipitate-matrix interface region was observed. The number of Frenkel pairs and their clustered fraction was not affected by the presence of the precipitate.     

Thermodynamic modelling of the C–U and B–U binary systems

The thermodynamic modelling of the carbon–uranium (C–U) and boron–uranium (B–U) binary systems is being performed in the framework of the development of a thermodynamic database for nuclear materials, for increasing the basic knowledge of key phenomena which may occur in the event of a severe accident in a nuclear power plant. Applications are foreseen in the nuclear safety field to the physico-chemical interaction modelling, on the one hand the in-vessel core degradation producing the corium (fuel, zircaloy, steel, control rods) and on the other hand the ex-vessel molten corium–concrete interaction (MCCI). The key O–U–Zr ternary system, previously modelled, allows us to describe the first interaction of the fuel with zircaloy cladding. Then, the three binary systems Fe–U, Cr–U and Ni–U were modelled as a preliminary work for modelling the O–U–Zr–Fe–Cr–Ni multicomponent system, allowing us to introduce the steel components in the corium. In the existing database (TDBCR, thermodynamic data base for corium), Ag and In were introduced for modelling AIC (silver–indium–cadmium) control rods which are used in French pressurized water reactors (PWR). Elsewhere, B₄C is also used for control rods. That is why it was agreed to extend in the next years the database with two new components, B and C. Such a work needs the thermodynamic modelling of all the binary and pseudo-binary sub-systems resulting from the combination of B, B₂O₃ and C with the major components of TDBCR, O–U–Zr–Fe–Cr–Ni–Ag–In–Ba–La–Ru–Sr–Al–Ca–Mg–Si + Ar–H. The critical assessment of the very numerous experimental information available for the C–U and B–U binary systems was performed by using a classical optimization procedure and the Scientific Group Thermodata Europe (SGTE). New optimized Gibbs energy parameters are given, and comparisons between calculated and experimental equilibrium phase diagrams or thermodynamic properties are presented. The self-consistency obtained is quite satisfactory.     

Zr-2/Cr扩散反应层物相分析

研究Zr-2/Cr扩散反应层物相，可为判断Zr-2和Cr是否相容提供依据。用热压法（50MPa)获得在1073K时Zr-2/Cr扩散反应层。分别用透射电镜（TEM）和场发射扫描电镜配备的薄窗能谱仪（EDS）对反应层进行结构的成分分析。结果表明，Zr-2/Cr扩散反应生成六方结构（C14型的）Zr(Fe,Cr)2Laves相。     

Irradiation-Induced Dissolution of Precipitates in Zircaloy-2

Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix. Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe. 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe. Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.     

Effect of Si and Zr on the interdiffusion of U-Mo alloy and Al

We investigated the effect of Zr additions to U-Mo and Si additions to Al on interdiffusion between U-Mo and Al by employing diffusion couple tests. We examined the phase stability of the γ-heat-treated alloys by high-temperature annealing tests. Using X-ray diffraction, we observed that the γ-phase U-7Mo-Zr alloys with more than 2 wt% Zr decomposed faster than the U-7Mo alloys. The diffusion couples showed that a Zr addition to U-7Mo and the addition of Si in Al reduced the interaction layer growth rates. However, Zr additions to U-Mo are most effective in reducing the overall interdiffusion rates when combined with Si additions to Al. The decomposition of the metastable U-Mo γ-phase during the diffusion test appears to have a significant effect on the overall interdiffusion rates.     

Angular Distributions of Resonance Neutron Scattering from Fe-56

Chromium depletion near grain boundaries of austenitic stainless steel during irradiation was investigated. Specimens were kept at 1,473 K for 30 min, and were quenched into the water. Irradiations were done using 400 keV He⁺ ions at 573, 673 and 773 K up to 10dpa with a dose rate of 2.4×10^?4 dpa/s. After irradiation, the Cr concentration profile near the grain boundary was measured using an analytical electron microscope with a 1 nm beam diameter. At 573 K, Cr depletion is small, and its concentration at the grain boundary decreases to 15.5 mass% at 3 dpa from the initial concentration of 18.5 mass%. At 673 and 773 K, Cr concentration at the grain boundary rapidly decreases between 0 and 0.2dpa, and then gradually approaches a constant value, 7.0 mass% at 673 K and 5.0 mass% at 773 K. Two stages are found in radiation induced segregation (RIS) behavior, one stage in which Cr depletion and Ni enrichment balance and another in which Fe depletion and Ni enrichment balance.

These experimental results were compared with the calculations based on the vacancy-induced inverse Kirkendall effect. Predicted Cr segregation at 673 and 773 K above 3dpa agrees with the experimental results. But Cr depletions at low doses which were obtained in the experiments are much faster than calculated. At 573 K in the experiments, depletion is smaller than calculated up to 10dpa.     

Segregation of alloying elements to free surfaces during irradiation

Recent transmission electron microscopy examinations of a number of face-centered-cubic and body-centered-cubic metals and alloys irradiated by heavy ions or by high-energy electrons have shown thatdynamic interactions of displacement damage with impurities and alloying elements lead to segregation and/or to the formation of second phases at internal surfaces such as voids. To date, the phenomenon has been observed in an experimental 18Cr8Ni1Si stainless steel, in commercial 316L stainless steel, in vanadium and in nickel. In the electron irradiated Fe18Cr8Ni1Si alloy, analysis of the segregation-induced strain field around the voids indicates that during irradiation minor substitutional alloying elements with negative and positive size factors segregate towards and away from the void surface respectively. Preliminary Auger spectroscopy analysis indicates that a similar segregation phenomenon occurs at the external irradiated surface in nickel-ion bombarded 18Cr8Ni1Si stainless steel. These results suggest that undersized substitutional elements may tend to preferentially interchange positions with oversized solutes in interstitial sites, and that transport by interstitials may dominate segregation to defect sinks.     

低锡Zr—4包壳管电子束焊接时发生的合金元素蒸发现象

采用电子探针的波谱分析方法,对国产低锡Ｚｒ－４包壳管的环焊缝试样进行表面成份分析。分析结果表明,从焊缝的外边到内边缘,Ｓｎ,Ｃｒ,Ｆｅ元素的化学成份在统计上呈增大趋势,腐蚀后出现了白色产物的试样表层,其Ｓｎ,Ｃｒ,Ｆｅ元素含量相当程度地降低。这一事实表明,国产低锡Ｚｒ－４包壳管采用电子束焊接时,在一定的焊接规范环焊缝的合金元素存在严重蒸发现象,特别是合金中锡元素的蒸发使其锡元素含量低于０．５％,导     

Residence Time of Crud on Surfaces of Channel Box in JPDR

Characterization of crud on surfaces of the channel box in JPDR has been carried out by means of chemical, radiochemical, X-ray diffraction and infrared spectrum analyses. The main cations in the crud are Fe and Ni: The sum of their weights amounts to more than 90% of the total weight of the cations found. The results of X-ray diffraction and infrared spectrum analyses revealed that the crud consisted of Ni_0.65Fe_2.35O₄, NiO and γ-FeOOH.

Based on the neutron flux calculated from the burn-up of ²³⁵U and ²³⁸U in the spent fuel, the apparent residence time of elements on the surface of the channel box was calculated to be 230 d for Co, 260 d for Ni and 70 d for Fe. The value for Fe should be taken as a minimum value, because of the presence of γ-FeOOH in the crud, which has been formed during the storage in a pond.

The present data are discussed in correlation with the one in the reactor water.