Annealing Temperature Effect on Creep Property of CZ Alloy for PWR Modern Fuel Claddings

2017 ◽  
Author(s):  
Yang Xu ◽  
Jun Tan ◽  
Liu-tao Chen ◽  
Dun-gu Wen ◽  
Hong Zou ◽  
...  
Keyword(s):  
Creep Rate ◽  
Creep Resistance ◽  
Nuclear Power ◽  
Annealing Temperature ◽  
Creep Property ◽  
Zirconium Alloys ◽  
Creep Stress ◽  
Lower Annealing Temperature ◽  
Different Temperatures ◽  
Water Reactors

Zirconium alloys remain as the main cladding materials in most water reactors. CZ zirconium alloy is recently developed for advanced PWR fuel assembly (STEP) designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 and CZ2 respectively. It is well known that creep property is one of the most important characteristics to evaluate the integrated performance of new commercial alloys. In this study, we have collected the creep date of CZ alloys heat-treated at different temperatures. The result shows that the creep rates of the zirconium alloys enlarge with the creep stress in the temperature of 375 °C. The creep resistance is improved with the increasing annealing temperature. The creep resistance of CZ-RXA (high temperature) is better than CZ-SRA (low temperature). The creep rate of CZ alloys at relatively lower annealing temperature is much larger than that at higher annealing temperature. At the given stress condition, the creep behaviors of different annealed CZ alloys are found to have almost the same tendency. In addition, the creep rate of CZ1 is smaller than CZ2 at the same annealing temperature.

Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys

2018 ◽  
Author(s):  
Lin Shi ◽  
Liutao Chen ◽  
Yang Xu ◽  
Changyuan Gao ◽  
Jun Tan ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Annealing Temperature ◽  
Creep Property ◽  
Pressurized Water Reactor ◽  
Lower Stress ◽  
Creep Tests ◽  
Pressurized Water ◽  
Final Annealing ◽  
Technology Institute ◽  
Cladding Material

Two kinds of new cladding material for pressurized water reactor were developed by China Nuclear Power Research and Technology Institute (CNPRI) which named CZ1 and CZ2. The cladding tubes of CZ alloys used in the study were fabricated by different final annealing temperature in the range of 450 °C to 600 °C.In order to investigate effect of final annealing temperature on creep property of CZ alloys. The axial creep tests were carried out on specimens of CZ-SRA and CZ-RXA At the temperature of 375 °C with different applied stress levels for 1000 hours. The experimental results showed that the axial creep resistance of CZ-RXA is superior to CZ-SRA at lower stress and is inferior to CZ-SRA at higher stress. The mechanism was discussed.

Investigation on High-Temperature Creep Properties of High-Alumina Bauxite

2016 ◽  
Vol 680 ◽  
pp. 347-351
Author(s):  
Shuo Cao ◽  
Yong Li ◽  
Jia Lin Sun ◽  
Hao Bo Zhang ◽  
Yong Qiang Sun ◽  
...  
Keyword(s):  
High Temperature ◽  
Liquid Phase ◽  
Creep Rate ◽  
Creep Resistance ◽  
Creep Property ◽  
High Alumina ◽  
High Temperature Creep ◽  
Creep Properties ◽  
Temperature Creep ◽  
Heat Preservation

This paper studied the high temperature creep properties of high-alumina bauxite (the mass fraction of Al2O3 in the new ore is about 78, the following abbreviations for Al2O3~78). The results indicated that the Al2O3~78 high-alumina bauxite mainly are corundum phase after high temperature sintered.When the temperature is 1100°C, corundum exists as crystal phase and the connections between grains are directly. The creep resistance of samples is very good at this temperature and the creep rate of 50 hours heat preservation is-0.266%. When the temperature is 1200°C, liquid phase starts to produce in a large number and the creep rate in 50 hours heat preservation is-1.589%. When the temperature is 1300°C, because of the further increase on the amount of liquid phase and wetting coated corundum particle, the direct connections between corundum particles are broken and the creep resistance is greatly reduced, the creep rate in 50 hours heat preservation is-4.088%. The creep curve fitting after 25 hours indicated that the creep property shows linear relations in three different temperatures after 25 hours. When the temperature is 1200°C and 1300°C, the creep variables arise rapidly in linear which declare the creep resistance of corundum is poor and increasing with temperature go up, more corundum phase is covered by glass phase and the creep resistance reduces dramatically.

scholarly journals Influence of Cu on the Improvement of Magnetic Properties and Structure of L10 FePt Nanoparticles

Nanomaterials ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 1097
Author(s):  
Luran Zhang ◽  
Xinchen Du ◽  
Hongjie Lu ◽  
Dandan Gao ◽  
Huan Liu ◽  
...  
Keyword(s):  
Magnetic Properties ◽  
Annealing Temperature ◽  
Reduction Method ◽  
Solid Phase ◽  
Average Particle Size ◽  
Fept Nanoparticles ◽  
Average Particle ◽  
Phase Reduction ◽  
Different Temperatures ◽  
One Step

L10 ordered FePt and FePtCu nanoparticles (NPs) with a good dispersion were successfully fabricated by a simple, green, one-step solid-phase reduction method. Fe (acac)3, Pt (acac)2, and CuO as the precursors were dispersed in NaCl and annealed at different temperatures with an H2-containing atmosphere. As the annealing temperature increased, the chemical order parameter (S), average particle size (D), coercivity (Hc), and saturation magnetization (Ms) of FePt and FePtCu NPs increased and the size distribution range of the particles became wider. The ordered degree, D, Hc, and Ms of FePt NPs were greatly improved by adding 5% Cu. The highest S, D, Hc, and Ms were obtained when FePtCu NPs annealed at 750 °C, which were 0.91, 4.87 nm, 12,200 Oe, and 23.38 emu/g, respectively. The structure and magnetic properties of FePt and FePtCu NPs at different annealing temperatures were investigated and the formation mechanism of FePt and FePtCu NPs were discussed in detail.

scholarly journals Solid state phase transformation kinetics in Zr-base alloys

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
A. R. Massih ◽  
Lars O. Jernkvist
Keyword(s):  
Phase Transformation ◽  
Solid State ◽  
Zirconium Alloys ◽  
Excess Oxygen ◽  
Transformation Kinetics ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Phase Transformation Kinetics ◽  
Solid State Phase Transformation ◽  
Water Reactors

AbstractWe present a kinetic model for solid state phase transformation ($$\alpha \rightleftharpoons \beta$$ α ⇌ β ) of common zirconium alloys used as fuel cladding material in light water reactors. The model computes the relative amounts of $$\beta$$ β or $$\alpha$$ α phase fraction as a function of time or temperature in the alloys. The model accounts for the influence of excess oxygen (due to oxidation) and hydrogen concentration (due to hydrogen pickup) on phase transformation kinetics. Two variants of the model denoted by A and B are presented. Model A is suitable for simulation of laboratory experiments in which the heating/cooling rate is constant and is prescribed. Model B is more generic. We compare the results of our model computations, for both A and B variants, with accessible experimental data reported in the literature covering heating/cooling rates of up to 100 K/s. The results of our comparison are satisfactory, especially for model A. Our model B is intended for implementation in fuel rod behavior computer programs, applicable to a reactor accident situation, in which the Zr-based fuel cladding may go through $$\alpha \rightleftharpoons \beta$$ α ⇌ β phase transformation.

scholarly journals Optimizing the Properties of La0.8Sr0.2CrO3 Thin Films through Post-Annealing for High-Temperature Sensing

Nanomaterials ◽  
2021 ◽  
Vol 11 (7) ◽  
pp. 1802
Author(s):  
Dan Liu ◽  
Peng Shi ◽  
Yantao Liu ◽  
Yijun Zhang ◽  
Bian Tian ◽  
...  
Keyword(s):  
Thin Film ◽  
Thin Films ◽  
Annealing Temperature ◽  
Rf Sputtering ◽  
Maximum Size ◽  
Cross Sectional ◽  
Post Annealing ◽  
Maximum Conductivity ◽  
Different Temperatures ◽  
Post Annealing Temperature

La0.8Sr0.2CrO3 (0.2LSCO) thin films were prepared via the RF sputtering method to fabricate thin-film thermocouples (TFTCs), and post-annealing processes were employed to optimize their properties to sense high temperatures. The XRD patterns of the 0.2LSCO thin films showed a pure phase, and their crystallinities increased with the post-annealing temperature from 800 °C to 1000 °C, while some impurity phases of Cr2O3 and SrCr2O7 were observed above 1000 °C. The surface images indicated that the grain size increased first and then decreased, and the maximum size was 0.71 μm at 1100 °C. The cross-sectional images showed that the thickness of the 0.2LSCO thin films decreased significantly above 1000 °C, which was mainly due to the evaporation of Sr2+ and Cr3+. At the same time, the maximum conductivity was achieved for the film annealed at 1000 °C, which was 6.25 × 10−2 S/cm. When the thin films post-annealed at different temperatures were coupled with Pt reference electrodes to form TFTCs, the trend of output voltage to first increase and then decrease was observed, and the maximum average Seebeck coefficient of 167.8 µV/°C was obtained for the 0.2LSCO thin film post-annealed at 1100 °C. Through post-annealing optimization, the best post-annealing temperature was 1000 °C, which made the 0.2LSCO thin film more stable to monitor the temperatures of turbine engines for a long period of time.

scholarly journals Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Coatings ◽  
2021 ◽  
Vol 11 (5) ◽  
pp. 557
Author(s):  
Egor Kashkarov ◽  
Bright Afornu ◽  
Dmitrii Sidelev ◽  
Maksim Krinitcyn ◽  
Veronica Gouws ◽  
...  
Keyword(s):  
Protective Coatings ◽  
Zirconium Alloys ◽  
Nuclear Industry ◽  
Normal Operation ◽  
Severe Accidents ◽  
Recent Developments ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Directed Research

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

14C Distribution in Atmospheric and Aquatic Environments Around Qinshan Nuclear Power Plant, China

Radiocarbon ◽  
2014 ◽  
Vol 56 (3) ◽  
pp. 1107-1114 ◽  
Author(s):  
Zhongtang Wang ◽  
Dan Hu ◽  
Hong Xu ◽  
Qiuju Guo
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Samples ◽  
Nuclear Power ◽  
Dissolved Inorganic Carbon ◽  
Tap Water ◽  
Surface Seawater ◽  
Pressurized Water ◽  
Specific Activities ◽  
Water Reactors

Atmospheric CO2 and aquatic water samples were analyzed to evaluate the environmental 14C enrichment due to operation of the Qinshan nuclear power plant (NPP), where two heavy-water reactors and five pressurized-water reactors are employed. Elevated 14C-specific activities (2–26.7 Bq/kg C) were observed in the short-term air samples collected within a 5-km radius, while samples over 5 km were close to background levels. The 14C-specific activities of dissolved inorganic carbon (DIC) in the surface seawater samples ranged from 196.8 to 206.5 Bq/kg C (average 203.4 Bq/kg C), which are close to the background value. No elevated 14C level in surface seawater was found after 20 years of operation of Qinshan NPP, indicating that the 14C discharged was well diffused. The results of the freshwater samples show that excess 14C-specific activity (average 17.1 Bq/kg C) was found in surface water and well water samples, while no obvious 14C increase was found in drinking water (groundwater and tap water) compared to the background level.

Finite Element Analyses of Residual Stresses in Typical Welded Joints Used in Nuclear Power Plants and Comparisons With Experiments

2010 ◽  
Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito
Keyword(s):  
Numerical Simulation ◽  
Residual Stress ◽  
Residual Stresses ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Butt Joint ◽  
Welding Residual Stress ◽  
Dissimilar Metal ◽  
Water Reactors

Recent discoveries of stress corrosion cracking (SCC) at nickel-based metals in pressurized water reactors (PWRs) and boiling water reactors (BWRs) have raised concerns about safety and integrity of plant components. It has been recognized that welding residual stress is an important factor causing the issue of SCC in a weldment. In this study, both numerical simulation technology and experimental method were employed to investigate the characteristics of welding residual stress distribution in several typical welded joints, which are used in nuclear power plants. These joints include a thick plate butt-welded Alloy 600 joint, a dissimilar metal J-groove set-in joint and a dissimilar metal girth-butt joint. First of all, numerical simulation technology was used to predict welding residual stresses in these three joints, and the influence of heat source model on welding residual stress was examined. Meanwhile, the influence of other thermal processes such as cladding, buttering and heat treatment on the final residual stresses in the dissimilar metal girth-butt joint was also clarified. Secondly, we also measured the residual stresses in three corresponding mock-ups. Finally, the comparisons of the simulation results and the measured data have shed light on how to effectively simulate welding residual stress in these typical joints.

scholarly journals 14C Release in Various Chemical Forms With Gaseous Effluents From the Paks Nuclear Power Plant

Radiocarbon ◽  
1989 ◽  
Vol 31 (03) ◽  
pp. 754-761 ◽  
Author(s):  
Ede Hertelendi ◽  
György Uchrin ◽  
Peter Ormai
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Liquid Scintillation ◽  
Mean Value ◽  
Scintillation Counting ◽  
Pressurized Water ◽  
The Mean ◽  
Water Reactors

We present results of airborne 14C emission measurements from the Paks PWR nuclear power plant. Long-term release of 14C in the form of carbon dioxide or carbon monoxide and hydrocarbons were simultaneously measured. The results of internal gas-proportional and liquid scintillation counting agree well with theoretical assessments of 14C releases from pressurized water reactors. The mean value of the 14C concentration in discharged air is 130Bqm-3 and the normalized release is equal to 740GBq/GWe · yr. > 95% of 14C released is in the form of hydrocarbons, ca 4% is apportioned to CO2, and <1% to CO. Tree-ring measurements were also made and indicated a minute increase of 14C content in the vicinity of the nuclear power plant.

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