Neutron irradiation of FeMn,FeCrMn and FeCrNi alloys and an explanation of their differences in swelling behavior

The swelling and phase stability of neutron-irradiated FeCrMn and FeCrNi alloys are compared in the range 420–600°C. While the behavior of the two alloy systems exhibits many similarities, that of the FeCrMn system is more complex, involving a higher level of phase instabilities. However, the sensitivity of radiation-induced density changes to composition is less in the FeCrMn system. In the FeCrMn system there are three major components of density change, namely void growth, ferrite formation and lattice parameter changes of the retained austenite; whereas void swelling is the only major component of density change in the FeCrNi system.     

Effects of Fe and Cr on corrosion behavior of ZrFeCr alloys in 500 °C steam

A study of the corrosion behaviors of ZrFeCr alloy and the influence of microstructure on corrosion resistance are described by X-ray diffraction and scanning electron microscope in this paper. The results show that several ZrFeCr alloys exhibit protective behavior throughout the test and oxide growth is stable and protective. The best alloy has the composition Zr1.0Fe0.6Cr. Fitting of the weight gain curves for the protective oxide alloys in the region of protective behavior, it showed nearly cubic behavior for the most protective alloys. The Zr1.0Fe0.6Cr has the more laves Zr(Fe,Cr)₂ precipitate in matrix and it has the better corrosion resistance. The Zr0.2Fe0.1Cr has little precipitate, the biggest hydrogen absorption and the worst corrosion resistance. The number of precipitates and the amount of hydrogen absorption in Zr alloy plays an important role on corrosion resistance behaviors in 500 °C/10.3 MPa steam.     

Evolution of nanoscopic iron clusters in irradiated zirconium alloys with different iron contents

We investigate the distribution of alloying elements in irradiated Zr alloys with different Fe contents using atom probe tomography. Our results showed dense nanoscale regions (clusters) of Fe formed in the matrix. The average diameter of the Fe clusters in alloy with a high Fe content increased under a higher neutron fluence. Conversely, the number density of Fe clusters remained similar in all the Zr alloy specimens. Energy-dispersive X-ray spectrometry showed that the maximum cluster size depended on the Fe content in the secondary phase particles. Fe clusters gathered along the basal plane of the Zr alloys at high fluence, indicating that irradiation defects influence Fe clustering. The solute concentration of Fe was estimated to be approximately 0.1 at%, which is the Fe concentration in the matrix exclusive of Fe clusters.     

铁基非晶态因瓦合金的热中子散射研究

因瓦（ｌｎｖａｒ）合金具有热膨胀性能随温度变化反常的特性，很多非晶态合金也具有显著的因瓦效应。用中子非弹性散射方法测量了铁基非晶态金属－金属型Ｆｅ９０－ｘＣｏｘＺｒ１０（ｘ＝１０，４０）和金属－类金属型Ｆｅ８０－ｙＣｒｙＰ１３Ｃ７（ｙ＝４，８）合金，以及Ｆｅ８６Ｃｏ４Ｚｒ１０因瓦合金居里点（ＴＣ＝３３０Ｋ）上下的广义声子谱。从３组声子谱的比较中，观察到了与因瓦效应相关的声子软化现象。结合已经观察到的晶态因瓦合金的声子软化现象，认为声子软化现象是因瓦效应的基本特征之一，它与晶格的长程有序无关，与材料类型无关。它的出现可能与在因瓦合金中存在增强的电子－声子相互作用有关     

A study of the structure of Zr(CrFe)2 laves phase precipitates in a ZrCrFe alloy

C15 type Zr(CrFe)₂ Laves phase precipitates have been found in Zr-1.15 wt% Cr-0.1 wt% Fe alloy. Twinned, multiple twinned and dislocation structures have been found in the precipitates. Comparison of calculated and measured precipitate size show the growth of the Zr(CrFe)₂ Laves phase is controlled by diffusion.The orientation relationships (1̄11̄)_L//(112̄0)_α, [110]_L//[0001]_α between the Zr(CrFe)₂ Laves phase precipitates and ga-Zr matrix in the ZrCrFe alloy give the same type of model for the transformation as previously suggested for Zircaloy-4.     

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.     

The role of Fe on the solubility of Nb in α-Zr

The solid solubility of Nb in α-Zr is an important parameter that has a potential impact on the corrosion properties of Zr-Nb alloys at reactor operating temperatures, i.e. below the monotectoid temperature. Work on dilute Zr-Nb alloys has shown that Fe is a common impurity that confounds the assessment of the solid solubility limit for Nb in Zr. This is because Fe has a very low solubility limit and it forms precipitates with both Nb and Zr. To assess the effect of Fe on the phases formed in the binary Zr-Nb alloy system, alloys containing 0.1-0.7 wt% Nb and <11 to 470 wt ppm Fe were heat-treated at temperatures between 575 °C and 600 °C and examined by transmission electron microscopy. Results indicate that, even at a concentration ? 24 ppm, Fe readily combines with Nb to form precipitates in the alloys with Nb contents in the range of 0.20 to 0.29 wt%. However, β-Nb particles were not observed for these same alloys and were only seen when the Nb content was ? 0.49 wt%. Because β-Nb particles were not found in the 0.29 wt% Nb alloy and the precipitation was estimated to have a negligible effect on the amount of Nb remaining in solution (reduced by <0.001 wt%), it is proposed that the solubility limit of Nb in a true binary Zr-Nb alloy would be between 0.29 and 0.49 wt%.     

Thermophysical properties of molten core materials: Zr–Fe alloys measured by electrostatic levitation

In order to investigate the progression of a core meltdown accident, it is necessary to understand the behavior of molten core materials. Zr–Fe alloys are one of the low-melting-temperature liquid phases that are thought to form in the early stages of bundle degradation. The objective of this study is to measure the thermophysical properties of Zr–Fe liquid alloys. Alloy samples with a composition of Zr_1?xFe_x (x = 0.12, 0.24, and 0.50) were synthesized by arc melting, and their density, viscosity, and surface tension were measured using an electrostatic levitation technique. The results indicate that the density of Zr–Fe liquid alloys can be estimated by a linear combination of the measured or extrapolated densities of pure Zr and Fe. The viscosities of the Zr–Fe liquid alloys can be roughly estimated by extrapolating those of Zr to lower temperatures, although this method tends to underestimate the viscosity of alloys, especially for eutectic compositions. The values of the Zr–Fe liquid alloys’ surface tensions are close to those of pure Zr.     

Nb、Cu对新型高温锆合金α/β相变温度和腐蚀性能的影响

研究不同元素含量的Zr-Nb-Cu合金的显微组织和其在500℃、10.3 MPa过热蒸汽中的耐腐蚀性能,结果表明,在500℃、10.3 MPa过热蒸汽中,Zr- 1.0Nb-0.05Cu合金的耐腐蚀性能最好,其耐腐蚀性能远远优于Zr-4和N18合金.在Zr-Nb-Cu合金中形成富含Nb、Fe、Cr的第二相粒子,这是影响锆合金耐腐蚀性能的一个原因.Zr-Nb-Cu合金在差热扫描量热仪分析的升温过程中,腐蚀产生的氢化物溶解,温度达到氢致α/β相变温度(约550℃)时开始β相变.添加Nb可以降低合金发生氢致β相变的温度,而增加Cu含量,可以降低合金腐蚀时的吸氢量,同时也使合金的耐腐蚀性能得到明显的提高.     

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.     

The improvement of irradiation-enhanced copper embrittlement in FeCu alloys

The effect of alloying elements on neutron irradiated FeCu alloys has been investigated in order to obtain the fundamental information on the irradiation-enhanced copper embrittlement for power reactor vessel steels. The mechanism of copper-induced irradiation embrittlement in the copper-containing iron alloys was proved to be due to both the interaction of copper atoms with irradiation-produced complex defects within grains, and the preferred grain boundary segregation of copper atoms existing near grain boundaries. The former effect causes the increase of yield strength, and the latter results in the ductility loss and grain boundary crackings. The addition of titanium up to 0.4 wt% to the Fe-0.1 wt% Cu alloy was found to be extremely effective in the improvement of both the irradiation-induced ductility loss and strength. Aluminum and silicon were not as effective as titanium.     

Mechanical properties and microstructure in neutron-irradiated nickel-based alloys and stainless steels for supercritical water-cooled-reactor fuel cladding

ABSTRACT

To investigate the irradiation behavior of mechanical properties and microstructural changes of commercial Ni-based alloys and improved stainless steels, a neutron-irradiation experiment was performed at the Joyo reactor, and post-irradiation examinations with tensile tests and TEM observations were carried out. The room-temperature tensile tests showed that all specimens that were irradiated at 485°C exhibited significant hardening and ductile behavior, especially in alloy 625. The irradiation hardening of all specimens irradiated at 668°C was less than that of specimens irradiated at 485°C. The fine-grained stainless steel, T3 and the Zr-added stainless steels, H1 and H2 showed good mechanical-property performance with keeping ductility after neutron irradiation. Most alloys and steels showed ductile behavior on the fracture surface except for alloy 625 specimen. The TEM observations showed that a high density of tangled dislocations and irradiation-induced defect clusters formed in the stainless steels and Ni-based alloys irradiated at 485°C. At 668°C, the material microstructures coarsened and their dislocation density decreased significantly. Long rod-like precipitates of Zr(Cr, Fe) compounds formed in the H1 and H2 steels that were modified with Zr. The yield stress drop of T3 steel in tensile stress was observed and is caused by grain-size coarsening at an irradiation of 668°C.     

Reactions of Uranium-Plutonium Alloys with Iron

In metallic U-Pu-Zr fuel for fast reactors, metallurgical reactions occur between the fuel alloy and the stainless steel cladding, and a liquid phase may be formed in the reaction zone at a higher temperature. In order to clarify the condition for liquefaction at the fuel-cladding interface, the reactions of U-Pu alloys with Fe have been examined at 923 and 943 K. The test results confirmed that the liquid phase is not formed at 923 K in any region of the reaction zone when the maximum Pu content in the (U,Pu)₆Fe phase is less than the Pu solubility limit in this phase. Comparison of the present test results with the liquefaction data from the various tests on metallic fuel-cladding compatibility suggested that the liquefaction condition is independent of the Zr content in the fuel alloy and can be expressed as a function of the atom fraction ratio of Pu/(U+Pu) in the fuel alloy and the reaction temperature. At 923 K, liquefaction will occur when the Pu/(U+Pu) ratio is larger than 0.25.     

Evolution of microstructure and second-phase particles in Zr–Sn–Nb–Fe alloy tube during Pilger processing

Evolution of microstructure and second-phase particles (SPPs) in Zr–Sn–Nb–Fe alloy tube were investigated during Pilger process using electron backscatter diffraction, secondary electron and transmission electron microscopy imaging techniques. Results show that the Pilger rolled tubes present heterogeneous structures with the C axes of less deformed grains mostly concentrated in the axial direction. During the Pilger rolling, the increase of deformation caused weakening of linear distribution of second-phase particles. The mean diameters of the precipitates are in the range of 70–100 nm in all specimens, and the growth mechanism of SPPs follows second-order kinetics. The grain growth is controlled by Zener pinning in the Pilger rolling–annealing specimens. Clusters containing the Zr(Nb,Fe)₂ and β_Nb precipitates formed in the Zr–1.0Sn–1.0Nb–0.12Fe alloy. Most of the particles located in grain boundaries are the Zr(Nb,Fe)₂ Laves phase with hexagonal structure, and stacking faults have been found in the Zr(Nb,Fe)₂ precipitates. The types, morphology and distribution of precipitates depend on the constituent and structural fluctuations of the nucleation area.     

Atomistic modeling of the interaction of cladding elements (Fe, Ni, Cr) with U-Zr fuel

Atomistic simulations of U-Zr fuel and its interaction with Fe, Ni, and Cr using the BFS method for alloys are presented. Results for the γU-βZr solid solution are discussed, including the behavior of the lattice parameter and coefficient of thermal expansion as a function of concentration and temperature. Output from these calculations is used to study the surface structure of γU-βZr for different crystallographic orientations, determining the concentration profiles, surface energy, and segregation behavior. The analysis is completed with simulations of the deposition of Fe, Ni and Cr on U-Zr substrates with varying Zr concentration. All results are discussed and interpreted by means of the concepts of strain and chemical energy underlying the BFS method, thus obtaining a simple explanation for the observed Zr segregation and its influence in allowing for cladding elements diffusion into the U-Zr fuel.     

Zr-0.2Cu-xNb合金的显微组织及腐蚀行为

对Zr-0.2Cu-x Nb(质量分数x=0.2,0.5,1.0,2.5)合金进行真空β相油淬、冷轧及退火处理,并在静态高压釜中进行过热蒸汽腐蚀试验,最后采用扫描电镜和透射电镜研究了合金及其腐蚀生成的氧化膜的显微组织。结果表明,随着Nb含量的增加,Zr-0.2Cu-x Nb合金中Zr2Cu第二相的数量逐渐减少,而β-Zr第二相数量逐渐增加;合金中尺寸较小的Zr2Cu第二相对耐腐蚀性能有利;β-Zr第二相在氧化过程中会促进氧化膜微裂纹的产生,降低合金的耐腐蚀性能。Zr-0.2Cu-x Nb合金中Nb含量接近其在α-Zr中最大固溶度时,合金具有最优的耐腐蚀性能。     

Effect of Cr on the creep properties of zirconium alloys

To investigate the effect of the Cr element in zirconium-based alloys on the creep properties, Zr–1.2Nb–0.1Cr, Zr–1.2Nb–0.5Cr, Zr–1.2Nb–0.3Sn–0.1Cr, and Zr–1.2Nb–0.2Sn–0.3Cr alloys were manufactured and creep tested under a constant stress of 120 MPa at 380 °C for 250 days. As the amount of Cr as well as Sn increased in the studied alloys, the creep strain rates decreased. The strengthening effect of Cr is considered to be efficient when the zirconium alloy contains Nb as an alloying element. The relative contribution of Cr against Sn contents on creep resistance was also observed to be comparable.     

Irradiated Behavior at High Burnup for HiFi Alloy

Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film.     

Impact of Material Configuration Behind First Wall on Helium Formation in In-Vessel Components of Fusion Reactor

The results of neutron transport calculations of the He formation based on the JENDL gas-production cross section file are discussed for some metals and alloys, namely ²⁷A1, Ti, ⁵¹V, Cr, ⁵⁵Mn, Fe, Ni, Zr, Mo, austenitic stainless steel (Ti modified 316 SS: PCA), Ni-base alloy (Inconel 625), ferritic steel (Fe-11Cr-1Mo: HT-9), Ti-base alloy (Ti-6A1-4V) and V-base alloy (V-5Cr-5Ti). Impacts of the two shields having the steel-rich and the H₂O-rich compositions and the two blankets having the Li₂O/Be-base and the liquid Li/Be-base compositions on the He formation rate in the above-mentioned metals and alloys are discussed. The relation between the He formation rate and the fast neutron flux (14.1 MeV>E>0.1 MeV) is investigated. The decrease of He formation at any distance Δ from the first wall more than Δ_as, the distance where the shape of neutron spectrum reaches its asymptotic form, is modelled by the simple formula based on the exponential dependence, as those reported so far for the fast neutron flux and the displacement damage rate.     

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