Specification for carbon and low alloy steel pressure vessels for primary circuits of nuclear reactors

2015 ◽  
Author(s):  
E. Kiss

To achieve high reliability for new designs and extended operation of Reactor Pressure Vessels and Internals it is mandatory to apply the technical knowledge gained during operation of the existing Plants to assure that sufficient “Margin” is built into the new design. This paper discusses the importance of four key structural degradation mechanisms that have been shown by operational experience to affect the reliability of the BWR. These are: 1) Stress Corrosion Cracking (IGSCC) of Stainless Steel and Nickel-based Alloys; 2) Irradiation Assisted SCC (IASCC) of Stainless Steel and Nickel-based Alloys; 3) Irradiation Embrittlement of RPV low alloy Steel; 4) Corrosion Assisted Fatigue of Carbon and Low Alloy Steel. While the focus of this paper is the BWR, the mechanisms discussed are equally applicable to the PWR, although the water chemistry effects and mitigations will be different.


2015 ◽  
Vol 1096 ◽  
pp. 27-30
Author(s):  
Wang Chen ◽  
Chen Jin ◽  
Yin Pei Wang

In chemical processing plants and petroleum refineries, the pressure vessels and pipelines suffered often from fire accidents and thus resulted in the metal materials were in overheating state. Elevated temperature could cause the changes in metallographic structure and have unfavorable effects on material mechanical properties. In order to understand better the influence laws of overheating on metallographic structures and material mechanical properties, the methods of theoretical analysis and experimental research were used and the effects of thermal exposure temperature, duration time and cooling rate on microstructure of low-alloy steel 12MnNiVR, were studied systematically. The study can provide basis data for the material properties database of metal material suffered from fire accident, and afford technical supports in the key technologies on fire damage FFS (Fitness-For-Service) integrity assessment.


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