scholarly journals Effect of Al Concentration on Microstructure and Properties of AlNbTiZr Medium-Entropy Alloy Coatings

Materials ◽  
2021 ◽  
Vol 14 (24) ◽  
pp. 7661
Author(s):  
Hongyang Xin ◽  
Jijun Yang ◽  
Jianjun Mao ◽  
Qingsong Chen ◽  
Jiaqi Yang ◽  
...  
Keyword(s):  
Corrosion Resistance ◽  
Composite Structures ◽  
Zirconium Alloy ◽  
Pressurized Water Reactor ◽  
Coating Materials ◽  
Pressurized Water ◽  
Surface Protection ◽  
Oxygen Penetration ◽  
Al Content ◽  
Operation Conditions

The AlNbTiZr medium-entropy alloy (MEA) coatings with different Al contents were prepared on N36 zirconium alloy substrates by RF magnetron co-sputtering. The morphology, microstructure, mechanical properties, surface wettability and corrosion resistance of the AlNbTiZr MEA coatings were studied to evaluate the surface protection behavior of zirconium alloy cladding under operation conditions of a pressurized water reactor. The results showed that all the coatings were composite structures with amorphous and bcc-structured nanocrystals. With the increase of Al content, both the elastic modulus and hardness decreased first and then increased. The hydrophobicity of the coatings was enhanced compared with that of the substrate. The 10.2 at.% Al AlNbTiZr coating had the best corrosion resistance and the minimum oxygen penetration depth, which originated from the formation of a denser oxide layer consisting of Nb2Zr6O17 and ZrO2. This study provides an improved idea for the design and development of Al-containing MEA coating materials for accident tolerant fuel.

JESSOP-SAVILLE ZIRCONIUM ALLOY

Alloy Digest ◽  
1965 ◽  
Vol 14 (3) ◽  
Keyword(s):  
Corrosion Resistance ◽  
Zirconium Alloy ◽  
Heat Treating ◽  
Pressurized Water Reactors ◽  
Neutron Absorption ◽  
High Melting Point ◽  
Absorption Properties ◽  
Pressurized Water ◽  
Excellent Corrosion Resistance ◽  
Water Reactors

Abstract JESSOP-SAVILLE ZIRCONIUM Alloy has a high melting point and possesses excellent corrosion resistance coupled with low neutron absorption properties. It is equivalent to ZIRCALOY 2. It is recommended for pressurized water reactors. This datasheet provides information on composition, physical properties, elasticity, and tensile properties as well as creep. It also includes information on corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: Zr-2. Producer or source: Jessop-Saville Ltd, Brightside Works.

scholarly journals Criticality safety analysis of spent fuel pool for a PWR using UO2, MOX, (Th-U)O2 and (TRU-Th)O2 fuels

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Claubia Pereira ◽  
Jéssica P. Achilles ◽  
Fabiano Cardoso ◽  
Victor F. Castro ◽  
Maria Auxiliadora F. Veloso
Keyword(s):  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Operation Conditions ◽  
Transport Code ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Criticality Safety

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized 

Radionuclide Composition of Airborne Discharges of European NPPS With WWER, Pressurized Water Reactor, and Boiling Water Reactor Facilities

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Angelina-Nataliya V. Vukolova ◽  
Andrei A. Rusinkevich
Keyword(s):  
Power Plants ◽  
Boiling Water ◽  
Normal Operation ◽  
Boiling Water Reactor ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Operation Conditions ◽  
Discharge Activity ◽  
Radionuclide Composition

Abstract The article presents the analysis of the data on radionuclide composition of airborne discharges of 52 European nuclear power plants (NPPs) with water–water energetic reactor facilities (WWER), pressurized water reactor facilities (PWR), and boiling water reactor facilities (BWR) under normal operation conditions. It contains lists of radionuclides, registered in discharges of researched NPPs, and gives estimation of contributions of radionuclides, forming the discharge, into total activity of discharge and into total effective dose, created by the discharge activity. It was determined that the maximal contribution into discharge activity of all researched NPPs make noble gases, tritium, and carbon-14, while the latter is the main dose-making radionuclide.

The Preliminary Performance Study of ATF With M3 Fuel and Zr-4 Cladding

2016 ◽  
Author(s):  
Yin Chunyu ◽  
Tu Teng ◽  
Changbing Tang ◽  
Yongjun Jiao
Keyword(s):  
Zirconium Alloy ◽  
Large Scale ◽  
Pressurized Water Reactor ◽  
Performance Study ◽  
Alloy Matrix ◽  
Pressurized Water ◽  
Research Directions ◽  
Silicon Carbon ◽  
Fuel Failure ◽  
Operation Parameters

Metal matrix microencapsulated (M3) fuel is one of the research directions on Accident Tolerant Fuel (ATF). In this article, it provides one of ATF design which consists of BISO (Bistructural ISOtropic) particles embedded in a zirconium alloy matrix, and the cladding coating with silicon carbon (SiC). The temperature distribution of the ATF element has been built, and then the center temperature has also been calculated based on the operation parameters of the large-scale pressurized-water reactor. Simultaneity, the several factors of fuel failure is preliminary analyzed and calculated, especially the pressure shell failure mechanism.

Effects of Annealing on the Corrosion Property of CZ1 Alloy

2017 ◽  
Author(s):  
Liutao Chen ◽  
Jun Tan ◽  
Changyuan Gao ◽  
Dungu Wen ◽  
Hong Zou ◽  
...  
Keyword(s):  
Corrosion Resistance ◽  
Nuclear Power ◽  
Underlying Mechanism ◽  
Corrosion Property ◽  
Second Phase ◽  
Pressurized Water Reactor ◽  
Corrosion Properties ◽  
Pressurized Water ◽  
Second Phase Particles ◽  
The Matrix

CZ alloys were developed by China General Nuclear Power Group (CGN) for pressurized water reactor. To improve the corrosion resistance, the impacts of annealing on corrosion of CZ1 alloy were studied. The fuel cladding tubes of CZ1 alloy was fabricated by different annealing processes, with intermediate annealing in the range of 560 °C to 700 °C. Then the corrosion properties of CZ1 alloy and low Tin Zr-4 specimens were investigated by autoclave tests. The matrix and second phase particles along with the oxide characteristics were analyzed. The experimental results showed that the corrosion resistance of CZ1 alloy is superior to that of low Tin Zr-4,. Besides, Lower corrosion weight gain is obtained in the cladding tubes with decreasing annealing temperature, which correlates well with the smaller size of second phase particles. The underlying mechanism was discussed. The results in this study will be used as a guide for the material processing design of CZ1 alloy.

A Technique for the Selection of the Fuel Pin Diameter for a Uranium/Zirconium Alloy-Fueled Pressurized Water Reactor

1992 ◽  
Vol 97 (1) ◽  
pp. 16-26 ◽  
Author(s):  
Dale B. Lancaster ◽  
Robert L. Marsh ◽  
Daniel B. Bullen ◽  
Holger Pfeifer ◽  
C. Steve Erwin ◽  
...  
Keyword(s):  
Zirconium Alloy ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Pin ◽  
Selection Of

Leaching Studies of Cement-Glass Package Containing Sodium Borate

1986 ◽  
Vol 84 ◽  
Author(s):  
Masahiro Okamoto ◽  
Koichi Chino ◽  
Tsutomu Baba ◽  
Tatsuo Izumida ◽  
Fumio Kawamura ◽  
...  
Keyword(s):  
Sodium Silicate ◽  
Liquid Waste ◽  
Sodium Borate ◽  
Pressurized Water Reactor ◽  
X Ray Diffraction ◽  
Adsorption Ability ◽  
Pressurized Water ◽  
X Ray ◽  
Solidification Reaction ◽  
Main Components

AbstractA new solidification technique using cement-glass, which is a mixture of sodium silicate, cement, additives, and initiator of the solidification reaction, was developed for sodium borate liquid waste generated from pressurized water reactor (PWR) plants. The cement-glass could solidify eight times as much sodium borate as cement could, because the solidifying reaction of the cement-glass is not hindered by borate ions.The reaction mechanism of sodium silicate and phosphoric silicate (initiator), the main components of cement-glass, was studied through X-ray diffraction and compressive strength measurements. It was found that three- dimensionally bonded silicon dioxide was produced by polymerization of the two silicates. The leaching ratio of cesium from the cement-glass package was one-tenth that of the cement one. This low value was attributed to a high cesium adsorption ability of the cement-glass and it could be theoretically predicted accordingly.

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