The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

Zr-Fe-Nb合金中Zr(Nb,Fe)2相在400 ℃/10.3 MPa过热蒸汽中的腐蚀行为

Zr-1.0Fe-1.0Nb合金经β相油淬、冷轧变形及580 ℃/5 h退火处理,在静态高压釜中进行400 ℃/10.3 MPa过热蒸汽腐蚀试验,利用带EDS的SEM和HRTEM对合金基体以及腐蚀生成的氧化膜显微组织进行分析。结果表明：合金中主要存在正交的Zr 3Fe和密排六方的Zr(Nb,Fe)₂第二相。Zr(Nb,Fe)₂相在氧化过程中先转变成非晶组织,非晶进一步氧化转化为m-Nb₂O₅和m-Fe₂O₃相纳米晶态氧化物,最后扩散流失到腐蚀介质中;Zr(Nb,Fe)₂相氧化后的Fe、Nb元素发生扩散流失,且Nb的流失速度大于Fe,合金元素的扩散流失在氧化膜中留下大量缺陷,促进氧化膜由柱状晶向等轴晶形态演化而不利于合金的耐腐蚀。     

Measurements of hydrogen permeation and absorption in zirconium oxide scales

The hydrogen absorption and the permeation behavior through the oxide layer formed on modified Zircaloy-4 (Zry-4) alloys were investigated. The modified Zry-4 was prepared by altering the chemical composition of standard Zry-4. The tin content of Zry-4 (1.5 wt%) was reduced to 0.5 wt%, and alloying elements Si, O and Nb were added from 0.01 to 0.2 wt%. The oxide layers were grown in a static autoclave at 360 °C under 18.3 MPa for 150 days. Fick’s law was used to calculate the diffusivity of hydrogen after the steady state of the permeation flux was reached. The diffusivity of hydrogen in the 0.5Sn–0.1Nb–0.1Fe–0.2Cr–0.2O–Zr specimen was lower than that in the 1.5Sn–0.2Fe–0.1Cr–0.1O–0.01Si–Zr and Zry-4 specimens. As the area fraction of precipitates increased, the hydrogen diffusivity increased whereas an inverse relationship between the diffusivity and the amount of the tetragonal phase was observed. In addition to the oxide structural study, the effects of the microstructure of the zirconium alloys such as precipitates and grain boundaries on the hydrogen absorption were studied.     

动力堆模拟高放废液的甲醛脱硝及脱硝过程中的沉淀行为北大核心CSCD

两步法玻璃固化工艺中,高放废液可通过化学脱硝达到降低酸度的目的,常用的化学脱硝剂有甲酸、甲醛、蔗糖等。以甲醛为化学脱硝剂,对动力堆模拟高放废液进行脱硝及脱硝过程中沉淀行为进行研究。模拟高放废液在90℃、脱硝比例为1.0~2.0范围内进行脱硝,对脱硝后各物质运用电感耦合等离子发射光谱(ICP-OES)、X射线荧光光谱(XRF)、扫描电镜-能谱分析(SEM-EDS)、Raman光谱、X射线衍射(XRD)、热重分析仪(TG)进行分析。结果表明:脱硝后废液中NO_(3)^(-)含量明显降低,随脱硝比例增大,NO_(3)^(-)的含量逐渐降低,甲醛含量增加。脱硝过程中出现由Zr、Mo、La、Ce、Nd、Fe、Te、Pr、Cs、Sm、Cr、Sr、Y、Co、Ni组成的沉淀,沉淀的形成有两个过程:一个过程为形成颗粒状的结晶物Ln_(2)Zr_(3)(MoO_(4))_(9)(Ln=La、Ce、Nd、Pr、Eu、Sm)和MoO_(2);另外一个过程为形成由O、Fe、Zr、Mo、Te构成的无定形粉末物质;脱硝产物的热分解主要发生在约360℃以下。     

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.     

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.     

Evaluation of effect of hydrogen on toughness of Zircaloy-2 by instrumented drop weight impact testing

Hydride-assisted degradation in fracture toughness of Zircaloy-2 was evaluated by carrying out instrumented drop-weight tests on curved Charpy specimens fabricated from virgin pressure tube. Samples were charged to 60 ppm and 225 ppm hydrogen. Ductile-to-brittle-transition behaviour was exhibited by as-received and hydrided samples. The onset of ductile-to-brittle-transition was at about 130 °C for hydrided samples, irrespective of their hydrogen content. Dynamic fracture toughness (K_ID) was estimated based on linear elastic fracture mechanics (LEFM) approach. For fractures occurring after general yielding, the fracture toughness was derived based on equivalent energy criterion. Results are supplemented with fractography. This simple procedure of impact testing appears to be promising for monitoring service-induced degradation in fracture toughness of pressure tubes.     

甲基膦酸二甲庚酯对Zr(Ⅳ)的萃取行为

以正十二烷为稀释剂,研究了甲基膦酸二甲庚酯(DMHMP)萃取剂对硝酸介质中Zr(Ⅳ)的萃取性能。从3.0 mol/L HNO₃中萃取Zr(Ⅳ)的分配比与萃取剂浓度及硝酸根浓度的关系表明：萃取过程中DMHMP以中性萃取剂形式与Zr(Ⅳ)配位,萃取反应方程式主要为： Zr⁴⁺+2DMHMP+4NO^-₃=Zr(NO₃)₄·2DMHMP 随着硝酸浓度的增大,还会生成Zr(NO₃)₄·2DMHMP·2HNO₃和Zr(NO₃)₄·2DMHMP·3HNO₃。该反应为放热反应,降低温度有利于DMHMP对Zr(Ⅳ)的萃取。     

基于W型试样确定A508-Ⅲ钢参考温度T0的方法研究

在我国早期开发的W型试样的基础上,参考ASTM E1921标准,开发了基于W型试样的断裂韧性测试技术,建立了包括断裂韧性计算、数据有效性判定和参考温度T₀计算等在内的数据分析方法。在-100～-40 ℃下开展了国产A508-Ⅲ钢的W型试样和标准1C(T)试样的测试分析和试验数据的有效性评价。结果表明,基于W型试样可得到满足ASTM E1921标准的有效断裂韧性数据,W型试样数据点均在标准1C(T)试样master curve(主曲线)的置信区间内,基于W型试样确定的参考温度T₀与标准1C(T)试样的非常接近,W型试样可成为RPV辐照监督备选试样。     

Influence of hydrogen content on fracture toughness of CWSR Zr–2.5Nb pressure tube alloy

In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr–2.5Nb pressure tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial–circumferential plane of the tube. Fracture toughness tests were carried out in the temperature range of 30–300 °C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (J_Q, J_Max and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (J_Q) and crack propagation (J_Max, and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value.     

Poisoning and saturation of St 737 getter alloy in the conversion of isotopic waters to isotopic hydrogen

The results of the studies performed to investigate the saturation capacity of St 737 getter alloy (Zr[V_0.5Fe_0.5]₂) in its application for conversion of water to hydrogen and the functional behaviour of the conversion reaction rate with large oxygen concentration in the alloy are reported. The experiment was performed with the getter alloy at a temperature of 400°C. The conversion process was monitored by two independent techniques — (i) by the hydrogen release, using a quadrupole mass spectrometer and (ii) by the quantity of water reduction in liquid phase. It was found that more than 100 mg of water was converted and 13 Pa m³ of hydrogen was released by a gram of alloy. A chemical analysis of the alloy performed after the experiment showed that the oxygen content in the alloy was about 7% by weight. The post-experiment diffraction analysis of the alloy showed that the crystalline structure of the alloy is almost completely destroyed and that the oxides formed are essentially amorphous. A brief discussion of the significance of the results for the application to tritiated water handling in future fusion reactions is also given.     

Experimental studies on steel corrosion in Pb-Bi with steam injection

Experimental studies on steel corrosion were performed in simulated PBWFR (Pb-Bi cooled direct contact boiling water fast reactor) coolant environment. Some candidate steels of high Cr contents were immersed in steam-injected liquid Pb-Bi pool to investigate how their Cr contents and oxygen potential in Pb-Bi or (P_H2/P_H2O) in the steam influence their corrosion behaviors at temperature range of the reactor operation. Test specimens were made from eight types of steel with Cr contents ranged from 8 to 18%. The experiments were conducted by exposing these specimens to Pb-Bi pool where steam was injected. (P_H2/P_H2O) ratios of the steam were employed as experimental parameter, ranged from < 3×10⁻⁷ to 1×10⁻⁵ to control oxygen potential of Pb-Bi. Exposure temperatures were studied of 400, 450 and 500°C. It was found that 12Cr steel (HCM12/HCM12A) was the most resistant to corrosion and therefore a candidate reactor material.     

Segregation effects on the corrosion behaviour of a phosphorus-doped AISI type 304L stainless steel

The effect of grain boundary (GB) segregation on intergranular stress corrosion cracking (IGSCC) in hot water environments at 150°C and 250°C was studied in a P-doped AISI type 304L stainless steel. The extent of segregation was measured by an exposure test in boiling 5 N HNO₃ + 8g/L K₂Cr₂O₇ solution as well as by a potentiostatic etch test at +1325 mV (SHE) in 5 N H₂SO₄ solution. Although GB segregation was detected in all the aged specimens, IGSCC was shown by only the specimens aged for 550°C/1000 h. The results suggest that it is the GB chromium depletion, rather than the segregation of phosphorus at the GBs, that controls IGSCC of stainless steels in the hot water environments studied.     

Measurement of the mechanical properties of thermally-shocked zirconia-based simulated inert matrix fuel

The Vickers micro-hardness (H_V) was measured by an indentation technique of simulated ZrO₂-based Inert Matrix Fuel (IMF) material with a composition of Er_0.07Y_0.10Ce_0.15Zr_0.68O_1.915 in two different densities on sintered specimens and specimens thermally shocked with the quenching temperature differences (ΔTs) between 473 and 1673 K and compared with those of simulated MOX, namely, U_0.92Ce_0.08O₂. The H_V values obtained for two IMF materials were found higher, ranging from 6.37 GPa to about 7.84 GPa, depending on ΔT and the sintered density, than those obtained for the simulated MOX which are quasi-constant in the same range of ΔT with a mean value of 6.37 GPa. The fracture toughness (K_IC) was calculated from the measured H_V and the crack length, and it was found to exhibit a slight increase with increasing ΔT, ranging between 1.4 and 2.0 MPa m^1/2, while that of simulated MOX specimen is ranging between 0.8 and 1.1 MPa m^1/2. The thermally shocked specimens were observed with an optical microscope and analyzed in terms of microstructural changes and cracking patterns.     

Influence of the crack-tip hydride concentration on the fracture toughness of Zircaloy-4

The influence of a hydrogen concentration gradient at the crack-tip and hydride platelet orientation on the fracture toughness, fracture mode and micromechanisms of a Zircaloy-4 commercial alloy was studied. Fracture toughness was measured on CT specimens and the analysis was performed in terms of J-integral resistance curves at temperatures ranging from 293 to 473 K. Fracture toughness results of specimens containing higher hydrides concentration near the crack-tip region, preferentially orientated in the crack plane, were compared to those obtained from specimens with a homogeneous hydrogen distribution and different platelet orientation; specimens were obtained by charging them in loaded and unloaded condition, respectively. Changes on both macroscopic and microscopic fracture behaviour were observed at temperatures ranging from 293 to 343 K, and the results show the relevance of both hydride concentration and platelet orientation. The existence of a ductile-to-brittle transition is discussed at the light of these new results.     

阻氚用Fe-Al渗铝层表面稳态相Al2O3膜生长机理研究

为阐释Fe-Al渗铝层表面暂态相Al₂O₃膜向稳态相α-Al₂O₃膜的转变机理,探索稳态相α-Al₂O₃膜制备的氧化工艺参数范围,采用掠入射角X射线衍射仪、辉光放电光谱仪、聚焦离子束、透射电镜等,结合热重分析对CLAM钢基体Fe-Al渗铝层在940～980 ℃、1 Pa～20 kPa参数下的氧化生长行为进行了深入表征与机理分析。研究结果表明,在1 Pa～20 kPa氧分压范围内Al₂O₃膜生长初期反应速率常数随着氧分压的升高而增大,而后期反应速率常数反而随之下降;采用掠入射角X射线衍射仪对3～180 min氧化不同时期表面Al₂O₃膜的相结构进行了掠入射角分析,推测Al2O3膜的生长经历了从氧化初期形成暂态相γ-Al₂O₃（15 min）→过渡态相α-（Al_0.948Cr_0.052）₂O₃（30 min）→稳态相α-Al₂O₃（120～180 min）的演变过程,最短相转变时间约60～90 min,连续Al₂O₃膜厚度约2 000 nm;同时,结合聚焦离子束对30、120 min形成的Al₂O₃膜表面进行了精确定向切割制样,并采用透射电镜选区电子衍射分析验证了相转变前Al₂O₃膜结构为过渡态相α-（Al_0.948Cr_0.052）₂O₃（113）,转变后为稳态相α-Al₂O₃（113）,证实了Cr作为第三组元促进暂态相向稳态相α-Al₂O₃的转变规律。     

Investigation of Stable Al2O3 Scale on Fe-Al Aluminized Coating for Tritium Barrier

It is well known that α-Al₂O₃ phase has stablility performance, high permeation reduction factor and good resistance performance in liquid LiPb, which is considered as the reference tritium barrier coating in future fusion reactor. In order to study the formation mechanism of stable α-Al₂O₃ scales on fusion structure material, the oxidation behavior of Fe-Al aluminized coating on China Low Activated Martensitic （CLAM） steel was investigated under the oxygen partial pressure from 1 Pa to 20 kPa at the temperature of 940-980 ℃. The Al₂O₃ scales were analyzed by thermogravimetric analysis meter, grazing angle X-ray diffractometer, glow discharge spectrometer, focused ion beam and transmission electron microscope. A single continuous Al₂O₃ scales with the maximum thickness of about 2 000 nm was formed on the diffusion Fe-Al aluminized layer. Thermogravimetric analysis results show that the higher oxidation rate constant is achieved while increasing the oxygen partial pressure, and then oxidation rate constant decreases. The phase transformation of Al₂O₃ scales on the surface of Fe-Al aluminized coating was studied during different oxidation time ranges from 3 min to 180 min. The metastable γ-Al₂O₃ and α-（Al0.948Cr0.052）₂O₃ phases is formed in the earlier oxidation process and finally transformed to stable α-Al₂O₃ phase. The features of the transient α-（Al0.948Cr0.052）₂O₃（113） and α-Al₂O₃（113） were detected by GXRD and then confirmed by focused ion beam and transmission electron microscope.     

Reactions between U–Zr alloys and Fe at 923 K

Interdiffusion experiments were carried out at 923 K with the diffusion couples consisting of U–23 at.% Zr/Fe and U–23 at.% Zr–1 at.% Ce/Fe. The reaction layer adjacent to the Fe was a single Zr-depleted UFe₂ phase. The phases in the reaction layers were estimated consistently with the calculated U–Zr–Fe ternary isotherm. The diffusion path obtained in this study was similar to that reported for the U–Pu–Zr/HT9-steel couple at 923 K, when those paths were expressed on the (U+Pu)–Zr–(Fe+Cr) composition triangle. The reaction layers grew in proportion to the square root of the annealing time. The addition of approximately 1 at.% of Ce to the U–23 at.% Zr alloy has little effect on the reaction between U–23 at.% Zr and Fe.     

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.     

IG-11石墨材料断裂韧性数值模拟研究

通过使用基于内聚力模型（CZM）的扩展有限元方法（XFEM）对单边切口梁的三点弯曲试验进行数值模拟来研究核级石墨IG-11断裂韧性的尺寸效应,试验和分析中考虑了试样整体尺寸和厚度变化,并对数值分析中的材料断裂参数进行了敏感性研究。模拟所得断裂韧性范围为0.90～1.10 MPa•m^1/2,这与试验所测得的0.82～1.27 MPa•m^1/2接近。模拟结果表明,材料断裂功对数值分析的影响较小,而材料断裂时的抗拉强度对数值分析的影响较大;另外,核石墨的断裂韧性（K_IC）存在明显的尺寸效应,随模拟试样整体尺寸的增大,断裂韧性增大,最终趋于一定值。这与现有文献中的尺寸效应模型所得到的预测值以及试验结果吻合得较好。但试样厚度则对K_IC的变化无明显影响。