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.

Prediction of Radiation Embrittlement for Highly Irradiated Reactor Pressure Vessel Steels

2009 ◽  
Author(s):  
Hiroshi Matsuzawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Test Results ◽  
Prediction Formula ◽  
Irradiation Embrittlement ◽  
Nuclear Plants ◽  
Domestic Plant ◽  
Test Reactor ◽  
Reactor Pressure

There are 53 (fifty-three) nuclear power plants (both PWR and BWR type) are now under operating in Japan, and the oldest plant has been operating more than thirty years. These plants will be operated until sixty years for operation periods, and will be verified the integrity for assessment of nuclear plants for every ten years in Japan. Reactor Pressure Vessels (RPVs) are required to evaluate the reduction of fracture toughness and the increase of the reference temperature in the transition region. As the operating period will be longer, the prediction for these material properties will be more important. Recently the domestic prediction formula of embrittlement was revised based on the database of domestic plant surveillance test results for thirty years olds as the JEAC4201-2007 [7]. The adequacy for this prediction formula using for sixty year periods is verified by use of the results of the material test reactors (MTRs), but the effects of the accelerated irradiation on embrittlement has not been clear now. So, JNES started the national project, called as “PRE” project on 2005 in order to investigate how flux influences on the ΔRTNDT. In this project the RPV materials irradiated in the actual PWR plant have been re-irradiated in the OECD/Halden test reactor by several different fluxes up to the high fluence region, and the microstructual change for these materials will be investigated in order to make clear the cause of the irradiation embrittlement. In this paper the overall scheme of this project and the summary of the updated results will be presented.

Mechanical Engineering Aspects of Concrete Reactor Pressure Vessels for Oldbury Nuclear Power Station

1965 ◽  
Vol 180 (1) ◽  
pp. 927-948
Author(s):  
R. W. Lakin
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Mechanical Engineering ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Integral Type ◽  
Power Station ◽  
The Core ◽  
Reactor Pressure

The use of prestrcssed concrete vessels to contain a nuclear reactor is not in itself novel, as the French in their G2 and G5 vessels at Marcoule had pioneered this form of construction, but the Oldbury vessels contained the first reactors of the integral type in which the core, boilers and gas circuit are contained within the same vessel. This type of reactor had been under consideration for some time by the author's company, and during the early part of 1960 a study had been completed which showed that this design was both feasible and economically attractive. The design formed the basis for the Oldbury Power Station, construction of which started in 1962.

Selection of candidate materials for reactor pressure vessels: Application of irradiation embrittlement prediction models and a stringency level methodology

2017 ◽  
Vol 233 (5) ◽  
pp. 965-976 ◽  
Author(s):  
Álvaro Rodríguez-Prieto ◽  
Ana María Camacho ◽  
Miguel Ángel Sebastián
Keyword(s):  
Chemical Composition ◽  
Western Europe ◽  
Prediction Models ◽  
Pressure Vessels ◽  
Fourth Generation ◽  
Irradiation Embrittlement ◽  
Water Reactors ◽  
Stringency Level ◽  
Reactor Pressure ◽  
Selection Of

The selection of materials for the reactor pressure vessel manufacturing is a complex process that involves great responsibility because small differences in chemical composition can adversely affect the manufacturing process and the in-service behavior of the material. Thus, it is recommendable to perform previous materials pre-selection stages based on the state-of-the-art knowledge, integrating research results with standardized requirements and using simplifier materials selection methodologies like the stringency level method. To address this issue, an evaluation of the influence of chemical composition on the shift of the ductile-to-brittle transition temperature has been performed using the most used and consolidated prediction models that are R.G. 1.99 Rev.2, NUREG/CR-6551, and ASTM E 900-02. A proposal of maximum limits for copper, nickel, and phosphorous to avoid irradiation embrittlement has been presented to carry out the process. The results have been analyzed by using the stringency level methodology to support the decision process. To this end, a materials data collection has been carried out to analyze the requirements described by 20 different specifications of materials from first to fourth generation of light water reactors, covering the main designs of pressurized reactors from Western Europe, North America, Japan, and Russia. It can be concluded that more recently developed materials exhibit more stringent requirements than earlier developed materials.

Probabilistic Fracture Mechanics Analysis for Degraded Reactor Pressure Vessel in Pressurized Water Reactor Nuclear Power Plant

2014 ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou
Keyword(s):  
Nuclear Power ◽  
Structural Reliability ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of PWR reactor pressure vessels over their long term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the United States is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Further, two overcooling transients of steam generator tube rupture and pressurized thermal shock addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR reactor pressure vessels and can be also referred as its regulatory basis in Taiwan.

Development of Engineering Codes of Practice for Corrosion Fatigue

1983 ◽  
Vol 105 (3) ◽  
pp. 255-262 ◽  
Author(s):  
P. M. Scott ◽  
B. Tomkins ◽  
A. J. E. Foreman
Keyword(s):  
Fatigue Failure ◽  
Corrosion Fatigue ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Low Alloy Steels ◽  
Aqueous Corrosion ◽  
Codes Of Practice ◽  
Alloy Steels ◽  
Service Inspection ◽  
Reactor Pressure

The recognition that time-dependent effects of aqueous corrosion can have a large influence on fatigue failure in carbon and low alloy steels for nuclear reactor pressure vessels is having an important influence on the continuing process of revision and improvement of design and inspection codes. The contrasting requirements of engineering codes for design purposes or for in-service inspection purposes are discussed. The critical assumptions in each are highlighted and illustrated with examples incorporating recent improvements in the mechanistic understanding of the development of corrosion fatigue failure. The consequences of adopting a crack-tolerant approach in the design phase, as distinct from the evaluation of defects found in service, are also critically examined.

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.

Assessment of the Continued Need for Independent Requirements on Transition Temperature and Upper Shelf Charpy Impact Toughness Metrics in Regulations Pertaining to Nuclear Reactor Pressure Vessels

2020 ◽  
Author(s):  
Mark Kirk ◽  
Masato Yamamoto ◽  
Marjorie Erickson
Keyword(s):  
Transition Temperature ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Ferritic Steels ◽  
Theoretical Knowledge ◽  
Charpy Impact Toughness ◽  
The Relationship ◽  
Reactor Pressure

Abstract The toughness requirements for the ferritic steels used to construct the primary pressure boundary of a nuclear power plant include both transition temperature metrics as well as upper-shelf metrics. These separate specifications for transition and upper shelf toughness find their origins in decisions made during the 1970s and 1980s, a time when there was much less empirical and theoretical knowledge concerning the relationship between these quantities. Currently, significant evidence exists to demonstrate a systematic relationship between transition and upper shelf toughness metrics for RPV-grade steels and weldments (e.g., the equations in draft Code Case N-830-1, empirical correlation between Charpy transition temperature and upper shelf metrics, etc.). This paper explores these relationships and demonstrates that, in many cases, the joint specification of transition temperature and upper shelf toughness values is redundant and, therefore, unnecessary.

An Assessment of Mechanical Properties of A508-3 Steel Used in Chinese Nuclear Reactor Pressure Vessels

2015 ◽  
Vol 137 (3) ◽  
Author(s):  
Meifang Yu ◽  
Y. J. Chao ◽  
Zhen Luo
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Construction Materials ◽  
Pressure Vessels ◽  
Under Construction ◽  
Reactor Pressure

China has very ambitious goals of expanding its commercial nuclear power by 30 GW within the decade and wishes to phase out fossil fuels emissions by 40–45% by 2020 (from 2005 levels). With over 50 new nuclear power plants under construction or planned and a design life of 60 years, any discussions on structural integrity become very timely. Although China adopted its nuclear technology from France or USA at present time, e.g., AP1000 of Westinghouse, the construction materials are primarily “Made in China.” Among all issues, both the accumulation of the knowledge base of the materials and structures used for the power plant and the technical capability of engineering personnel are imminent. This paper attempts to compile and assess the mechanical properties, Charpy V-notch impact energy, and fracture toughness of A508-3 steel used in Chinese nuclear reactor pressure vessels (RPVs). All data are collected from open literature and by no means complete. However, it provides a glimpse into how this domestically produced steel compares with western RPV steels such as USA A533B and Euro 20MnMoNi55.

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