Detection of Irradiation Embrittlement of Low-Alloy Steel for Nuclear Reactor Pressure Vessels Using a Probe Type Eddy Current Sensor

2009 ◽  
pp. 601-601-15
Author(s):  
N Maeda ◽  
A Yamaguchi ◽  
T Sugibayashi ◽  
K Kohno
Keyword(s):  
Alloy Steel ◽  
Eddy Current ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Current Sensor ◽  
Probe Type ◽  
Low Alloy Steel ◽  
Irradiation Embrittlement ◽  
Eddy Current Sensor ◽  
Reactor Pressure

Component Reliability Considerations for New Designs and Extended Operation of Boiling Water Reactor (BWRs)

2008 ◽  
Author(s):  
E. Kiss
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
High Reliability ◽  
Pressure Vessels ◽  
Operational Experience ◽  
Low Alloy Steel ◽  
Irradiation Embrittlement ◽  
Component Reliability ◽  
Extended Operation ◽  
Reactor Pressure

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.

Heavy Section Steel Plates for Reactor Pressure Vessels Produced by Basic Oxygen Process

1977 ◽  
Vol 99 (2) ◽  
pp. 314-321
Author(s):  
H. Nakao ◽  
R. Yamaba ◽  
S. Takaishi ◽  
H. Kunitake ◽  
S. Kanazawa
Keyword(s):  
Electric Furnace ◽  
Nuclear Reactor ◽  
Temper Embrittlement ◽  
Pressure Vessels ◽  
Steel Plates ◽  
Irradiation Embrittlement ◽  
Basic Oxygen ◽  
Furnace Process ◽  
Oxygen Process ◽  
Reactor Pressure

Low alloy steel plates with heavy sections for pressure vessels of direct desulfurization unit and nuclear reactor were produced by basic oxygen process instead of the conventional electric furnace process. There was a decrease of impurity and residual elements which increase susceptibility to neutron irradiation embrittlement and temper embrittlement. Productivity also increased by this process. Various properties of the plates thus manufactured were evaluated in comparison with those of electric furnace plates. It was found as a result that the basic oxygen process produces an improved notch toughness for use in nuclear reactor pressure vessels, and approximately the same levels of properties for application to desulfurization-purpose pressure vessels, compared with the electric furnace process in which BOP return scrap was also used especially.

scholarly journals Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

2021 ◽  
Vol 13 (19) ◽  
pp. 10510
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Ana María Camacho ◽  
Carlos Mendoza ◽  
John Kickhofel ◽  
Guglielmo Lomonaco
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Prediction Models ◽  
Pressure Vessels ◽  
Defect Size ◽  
Historical Evolution ◽  
Nuclear Power Generation ◽  
Irradiation Embrittlement ◽  
Service Inspection ◽  
Reactor Pressure

The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

ASTM A 533 GRADE B

Alloy Digest ◽  
1977 ◽  
Vol 26 (1) ◽  
Keyword(s):  
Fracture Toughness ◽  
Physical Properties ◽  
Tensile Properties ◽  
Nickel Alloy ◽  
Alloy Steel ◽  
Nuclear Reactor ◽  
Heat Treating ◽  
Pressure Vessels ◽  
Steel Mills ◽  
Reactor Pressure

Abstract ASTM A 533 Grade B is a manganese-molybdenum-nickel alloy steel for use in the quenched-and-tempered condition for plate for the construction of pressure vessels and other pressure equipment. It is used in 6 to 12-inch sections and is the most widely used steel for nuclear reactor pressure vessels. This datasheet provides information on composition, physical properties, elasticity, and tensile properties as well as fracture toughness. It also includes information on forming, heat treating, machining, and joining. Filing Code: SA-328. Producer or source: Alloy steel mills and foundries.

Inspection of Surface Rolled Sintered Gears using Eddy-Current Sensor

2013 ◽  
Vol 133 (5) ◽  
pp. 300-306
Author(s):  
Tsutomu Mizuno ◽  
Yuichi Asato ◽  
Sho Goto ◽  
Takashi Watanabe ◽  
Teruie Takemasu ◽  
...  
Keyword(s):  
Eddy Current ◽  
Current Sensor ◽  
Eddy Current Sensor

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

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