Review of the initiation and arrest temperatures for delayed hydride cracking in zirconium alloys

Based on Kim's delayed hydride cracking (DHC) model, this study reanalyzes the critical temperatures for DHC initiation and arrest in zirconium alloys that had previously been investigated with Puls DHC model. In an unratcheting thermal cycle where DHC crack tip hydrides were dissolved fully at the peak temperature, the DHC initiation was suppressed, which required a supercooling or ΔT from the terminal solid solubility for dissolution (TSSD) temperatures. At a hydrogen concentration as 7 ppm H, the DHC initiation temperatures coincided with the TSSD, which is in conflict with Puls' DHC model. In a ratcheting thermal cycle, where the hydrides precipitated at the DHC crack tip were not fully dissolved, the DHC initiation was enhanced, so as to require a lesser ΔT, compared to that of the unratcheting thermal cycle. Therefore, the DHC initiation temperatures are concluded to depend upon at what temperatures the hydrides can nucleate in the zirconium matrix with the supersaturated hydrogen concentration. The DHC arrest temperatures were governed by the critical supersaturated hydrogen concentration or ΔC regardless of the thermal cycle treatment, providing further supportive evidence that Kim's DHC model is feasible.     

N18锆合金氢致裂纹延迟开裂临界温度研究

研究了N18锆合金(Zr-1Sn-0.3Nb-0.3Fe-0.1Cr)发生氢致延迟开裂(DHC)临界最大开裂温度(Tc)和临界最小止裂温度(Th)随氢含量的变化规律;同时对裂纹尖端偏聚氢含量及静水应力和发生DHC的临界氢含量进行了理论分析,建立理论模型对临界温度进行理论计算.结果表明:N18合金发生氢致延迟开裂的临界温度介于相同氢含量下溶解固溶温度与析出固溶温度之间,且最大开裂温度小于最小止裂温度,计算的临界温度值与实验值相当吻合,因此该理论模型能够真实反映N18锆合金的氢致延迟开裂的物理过程.     

A theoretical and experimental study of hydrides in zirconium alloys

A method developed for computing the critical length and thickness of hydride plates formed in delayed hydride cracking (DHC) in zirconium alloys is considered. The model is based on analyzing the distribution of tensile stresses in the plane of a sharp normal tensile crack. The characteristics of hydrides formed due to DHC in reactor tubes produced from alloy Zr-2.5% Nb are determined experimentally. The results of the computation agree well with experimental data. __________ Translated from Metallovedenie i Termicheskaya Obrabotka Metallov, No. 4, pp. 5–9, April, 2006.     

Driving force for delayed Hydride cracking of zirconium alloys

The effect of hydrogen concentration on the delayed hydride cracking velocity (DHCV) and the threshold stress intensity factor, K_IH of a Zr-2.5Nb tube were examined at test temperatures ranging from 100 to 280°C by subjecting compact tension specimens with a hydrogen concentration of 12 to 100 ppm H to an overtemperature cycle. The DHCV and K_IH increased and decreased, respectively, with an increase in the supersaturated hydrogen concentration over the terminal solid solubility for dissolution (TSSD) or ΔC. They then leveled off to constant values at ΔC in excess of the ΔC_max corresponding to a difference of the terminal solid solubility of the hydrogen on cool-down and on heat-up. Further, intentional introduction of an undercooling by 0 to 40°C at the test temperature decreased the DHCV of the Zr-2.5Nb tube, indicating that ΔC between the bulk region and the crack tip governs the DHCV. A new DHC model is proposed where the driving force for DHC is the difference in the hydrogen concentration between the bulk region and the crack tip by preferentially nucleating the hydrides only at the crack tip under an applied tensile stress, due to a hysteresis in the TSS of hydrogen on heat-up and on cool-down. A supplementary experiment was conducted to validate the feasibility of the proposed DHC model.     

计数器控制技术在级进模中的应用实例

通过对散热片零件的工艺分析 ,确定了用级进模加工的工艺排样步骤为 :冲凸—冲孔—翻边—撕口弯形—切断 ,介绍了散热片级进模的模具结构、主要尺寸和位置精度的保证方法 ,以及计数器DHC -JDM1 5对模具运行时的控制过程和散热片级进模的工作原理。     

Radiation-Resistant Alloys of the Nickel-Chromium System

Properties of some alloys of the nickel-chromium system are studied with the aim of determining the possibility of their use as structural materials for nuclear reactors. It is shown that at some compositions such alloys form a structure ensuring high process and service properties under irradiation in boiling water reactors and pressurized water reactors. Commercial production of these alloys has begun. __________ Translated from Metallovedenie i Termicheskaya Obrabotka Metallov, No. 7, pp. 78 – 82, July, 2005.     

Morphology Engineering of Porous Media for Enhanced Solar Fuel and Power Production

The favorable and adjustable transport properties of porous media make them suitable components in reactors used for solar energy conversion and storage processes. The directed engineering of the porous media’s morphology can significantly improve the performance of these reactors. We used a multiscale approach to characterize the changes in performance of exemplary solar fuel processing and solar power production reactors incorporating porous media as multifunctional components. The method applied uses imaging-based direct numerical simulations and digital image processing in combination with volume averaging theory to characterize the transport in porous media. Two samples with varying morphology (fibrous vs. foam) and varying size range (mm vs. μm scale), each with porosity between 0.46 and 0.84, were characterized. The obtained effective transport properties were used in continuum-scale models to quantify the performance of reactors incorporating multifunctional porous media for solar fuel processing by photoelectrochemical water splitting or power production by solar thermal processes.     

Versatile Oxide Films Protect FeCrAl Alloys Under Normal Operation and Accident Conditions in Light Water Power Reactors

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO₂–zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10–22 Cr + 4–6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.     

Zirconium-Niobium Alloys for Core Elements of Pressurized Water Reactors

The main characteristics of niobium-bearing zirconium alloys used for fabricating fuel element claddings of pressurized water reactors are considered. It is shown that the high corrosion and radiation resistance of zirconium parts is provided by the chemical composition, structure, and phase composition of the alloys. The Zr – Nb alloys developed in Russia provide reliable operation of fuel elements and fuel rod arrays in active reactors and serve as a basis for new modifications.     

Space fission reactor structural materials: Choices past,present, and future

Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermoelectric generator and solar technologies. The use of fission reactors for space applications has been studied for over 50 years. Structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15–20 year reliability with no inspection or preventative maintenance. Many different structural materials have been proposed. While all of those proposed meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for fission reactors for space power.     

Creep Behavior of Hydrogenated Zirconium Alloys

Zirconium (Zr) alloys are the primary structural materials of most water reactors. Creep is considered to be one of the important degradation mechanisms of Zr alloys during reactor operating and repository conditions. Zr alloys pick up hydrogen (H₂) during their service from the coolant water. Hydrogen can be present in solid solution or precipitated hydride form in Zr alloys depending upon the temperature and concentration. This study reviews the effect of hydrogen on creep behavior of Zr alloys used in the water reactors.     

Thermal conductivity of UO<Subscript>2</Subscript> fuel: Predicting fuel performance from simulation

Recent progress in understanding the thermal-transport properties of UO₂ for fission reactors is reviewed from the perspective of computer simulations. A path to incorporating more accurate materials models into fuel performance codes is outlined. In particular, it is argued that a judiciously integrated program of atomic-level simulations and mesoscale simulations offers the possibility of both better predicting the thermal-transport properties of UO₂ in light-water reactors and enabling the assessment of the thermal performances of novel fuel systems for which extensive experimental databases are not available.     

An innovative gas-solid torbed reactor for the recycling industries

Gas-solid Torbed reactors have been developed for processing a wide range of materials. The reactors have facilitated several novel recycling projects in countries where they are used. One of the major advantages of these reactors when applied to the recycling industries is the compactness of the plant and its inherent ability to be scaled down and fully automated to better match the volume requirements of this sector. Authors’ Note: Torbed is a registered trademark of Torftech Ltd. V.I. Lakshmanan earned his Ph.D. in chemistry at Bombay University, India, in 1968. He is currently program director of process technologies with Ortech Corporation. C.E. Dodson earned his diploma in mechanical engineering at Polytechnic of the South Bank, London, in 1968. He is currently president of Torftech Ltd. Mr. Dodson is a member of TMS.     

Application of Plasma Technology in Iron and Steelmaking

The potential for the application of plasma technology in metal oxide reduction and in iron and steelmaking is outlined and discussed. Recent evolution and developments in the plasma-based reactors employed in the production of iron, steel, and ferroalloys have been reviewed; the current status is outlined in terms of process control, flexibility in the raw materials consumed, product quality, and energy conservation. The advantages and limitations of thermal plasma-based reactors have been critically outlined and their potential to seriously challenge the blast furnace/basic oxygen furnace steelmaking route is considered.     

Zr-4合金中氢化物析出长大的透射电镜原位研究

用透射电子显微镜拉伸试样台原位研究了应力、电子束辐照以及第二相对Zr-4合金中氢化物析出和长大的影响。结果表明,在拉应力作用下,裂纹易于沿氢化物扩展,并在裂尖垂直于拉应力方向析出新的氢化物。氢化物在拉应力诱发下的析出、开裂、再析出·····过程,导致了氢致延迟开裂。在较强的会聚电子束辐照下,Zr-4合金中的氢化物会分解,新的氢化物会围绕着附近的Zr(Fe,Cr)2第二相粒子析出,新析出的氢化物为面心立方结构的δ相。     

Progress in corrosion resistant materials for supercritical water reactors

Supercritical water reactors (SCWRs) are a kind of high-temperature, high-pressure water-cooled reactors that operate above the thermodynamic critical point of water (374 °C, 22.1 MPa). Corrosion and degradation of materials used in supercritical water environments are determined by several environment- and material-dependent factors. In particular, irradiation-induced changes in microstructure and microchemistry are major concerns in a nuclear reactor. Many structural materials including alloys and ceramics have been proposed for use as SCWR components or materials for applying protective coatings in SCWRs. Various surface modification processes are also explored to change the chemical composition and microstructure of the near surface regions. This article aims to provide an overview of recent materials developments for supercritical water reactors focusing mainly on the nuclear reactor applications. The emphasis is placed on the corrosion and degradation mechanisms and the selection criteria of materials. In addition, the development of new processes for surface modification of materials in SCWRs is also briefly reviewed. Finally, some perspectives on the direction of future research in this area are also outlined.     

ADISCUSSION ON THE PROBLEMS PRESENT IN IN SERVICE HYDROGENATION REACTORS

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Cladding and duct materials for advanced nuclear recycle reactors

The expanded use of nuclear energy without risk of nuclear weapons proliferation and with safe nuclear waste disposal is a primary goal of the Global Nuclear Energy Partnership (GNEP). To achieve that goal the GNEP is exploring advanced technologies for recycling spent nuclear fuel that do not separate pure plutonium, and advanced reactors that consume transuranic elements from recycled spent fuel. The GNEP’s objectives will place high demands on reactor clad and structural materials. This article discusses the materials requirements of the GNEP’s advanced nuclear recycle reactors program.     

Progress in the pyrochemical processing of spent nuclear fuels

Pyrochemical processes are being developed to recover actinides from spent fuels from light-water reactors and integral fast reactors. The transuranic elements from light-water reactors will be introduced into the integral fast reactor fuel cycle. To meet the requirements of that fuel cycle, transuranic elements are recovered as oxide free metal containing some fission products. This article discusses pyrochemical processes for recovering actinides from light-water reactor and integral fast reactor fuels and for treating the high-level wastes from these processes. The development status of these processes and the plans to demonstrate them using facilities at Experimental Breeder Reactor II are also described.     

Production of Nanoparticles in Thermal Plasmas: A Model Including Evaporation,Nucleation, Condensation,and Fractal Aggregation

In this work a coupled model for the production of nanoparticles in an inductively coupled plasma reactor is proposed. A Lagrangian approach is used to describe the evaporation of precursor particles and an Eulerian model accounting for particle nucleation, condensation, and fractal aggregation. The models of the precursor and nanoparticles are coupled with the magneto-hydrodynamic equations describing the plasma. The purpose of this study is to develop a model for the synthesis of particles in a thermal plasma reactor, which can be used to optimize industrial reactors. The growth of aggregates is considered by introducing a power law exponent D _f. Results are compared qualitatively and quantitatively with existing experimental data from plasma reactors at a relatively large laboratory scale. The results obtained from the model confirm the previously observed importance of the quench strategy in defining the morphology of the nanoparticles.