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.

scholarly journals Some Useful Mechanical Property Correlations for Nuclear Reactor Pressure Vessel Steels

2018 ◽  
Author(s):  
Randy K. Nanstad ◽  
William L. Server ◽  
Mikhail A. Sokolov ◽  
G. Robert Odette ◽  
Nathan Almirall
Keyword(s):  
Mechanical Property ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Rigorous Approach ◽  
Pressure Vessel Steels ◽  
Physically Based ◽  
Induced Changes ◽  
The Relationship ◽  
Reactor Pressure

The use of correlations is common in the research and development arena of the nuclear industry with the realization that some applications with direct implications to safety demand a more rigorous approach. Most correlations involve the relationship between two experimental properties, such as that between hardness and tensile strength. There are others that are much more complicated and are often designated models because they incorporate physically-based knowledge; examples of this are predictive correlations for irradiation-induced embrittlement of reactor pressure vessels (RPV). The objective of this paper is to collect and discuss many of the commonly used correlations for applications to nuclear RPVs. This paper identifies and discusses various correlations that relate easily measured properties to properties that are more difficult, more time consuming, or more expensive to measure. In the case of irradiated RPV materials, irradiation-induced changes in easily measured properties are related to the changes in those more difficult to measure. It is noted that recognition and understanding of the uncertainties associated with all correlations is highly important.

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.

Reconstitution of V-Notched Charpy Impact Specimen Made of Short Insert

2006 ◽  
Author(s):  
Toru Osaki ◽  
Hiroshi Matsuzawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Absorption Energy ◽  
Impact Specimen ◽  
Insert Length ◽  
Reactor Pressure

Reconstitution in this paper means to constitute the original size V-notched Charpy impact specimen, which is made of the irradiated insert cut out from broken piece and un-irradiated tabs welded to the insert. It is a promising technique to secure an adequate number of surveillance specimens for long-term operation of nuclear power plants. Every Japanese nuclear power plant has its own surveillance test program, and is operated considering its unique surveillance test results along with the general reduction tendency of fracture toughness. This practice should be continued and enhanced if possible, after the full use of originally installed specimens, because its fracture toughness is lower than before. Reconstitution of V-notched Charpy impact specimens to the original shape by using a short insert was studied. Charpy absorption energy is generally shifted by reconstitution, if the insert length is as short as 10 mm. Reconstitution with a short insert is necessary when the transverse property of the original specimen is required although only the longitudinal surveillance specimen is installed as in some early constructed reactor pressure vessels in Japan. This case is important when the reactor pressure vessel is suspected to be a so-called low upper shelf toughness reactor pressure vessel. The minimum required insert length to avoid affect on the specimen properties depends on the Charpy absorption energy of the insert and reconstitution weld condition. Correlation between Charpy absorption energy and plastic deformation size, and short time annealing properties of irradiated pressure vessel steels were investigated. A method to evaluate the minimum required insert length was proposed, which depends on the expected Charpy absorption energy and thermal transient during reconstitution. It was demonstrated that the reconstituted specimens of 10 mm-long irradiated inserts, whose upper shelf absorption energy was 69J and required insert length was 9.5mm, showed little shift of upper shelf absorption energy.

Neutron Embrittlement Aging Management of Nuclear Reactor Pressure Vessels

2003 ◽  
Author(s):  
William L. Server
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Temperature Shift ◽  
Environmental Parameters ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Vessel Material ◽  
Reactor Pressure

The management of neutron embrittlement of nuclear reactor pressure vessels involves monitoring of the changes in the fracture toughness of surveillance capsule specimens that closely approximate the actual reactor vessel material(s). The measurement of fracture toughness is currently performed in an indirect manner using Charpy V-notch impact specimens, although the direct measurement of fracture toughness is possible using the same small Charpy specimens fatigue precracked to produce acceptable fracture toughness three-point bend specimens. This paper first examines the current Charpy-based approach and the development of a recent embrittlement correlation that has been incorporated into ASTM E 900-02, “Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials.” This correlation provides the latest mechanistically-guided approach to assess the changes in transition temperature shift. This same correlation and mechanistic guidance can be used with measured fracture toughness data developed following ASTM E 1921-02 to account for differences in surveillance material versus actual vessel material. Additionally, environmental parameters such as fluence and temperature also can be adjusted between different irradiation facilities using this latest correlation. This paper focuses on the application of the new ASTM E 900-02 correlation to Charpy-based and fracture toughness-based measurements to develop the best predictive approach for assuring structural integrity of reactor vessel materials. Key technical issues important for extended vessel life also are discussed.

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.

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.

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.

Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event

2010 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Low Temperature ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Probability Of Failure ◽  
Pressure Vessels ◽  
Boiling Water ◽  
Power Station ◽  
Circumferential Welds ◽  
Water Reactor ◽  
Reactor Pressure

With the development of probabilistic fracture mechanics (PFM) methods in recent years, the risk-informed approach has gradually been used to evaluate the structural integrity and reliability of the reactor pressure vessels (RPV) in many countries. For boiling water reactor (BWR) pressure vessels, it has been demonstrated that it is not necessary to perform the inservice inspections of beltline circumferential welds to maintain the required safety margins because their probability of failure is orders of magnitude less than that of beltline vertical welds, thus may well reduce the associated substantial cost and person-rem exposure. In Taiwan, however, the inservice inspections of shell welds still have to be performed every ten years per ASME Boiler and Pressure Vessel Code, Section XI inspection requirements for a BWR type Chinshan nuclear power station. In this work, a very conservative PFM model of FAVOR code consistent with that USNRC used for regulation is built with the plant specific parameters concerning the beltline shell welds of RPVs of Chinshan nuclear power station. Meanwhile, a hypothetical transient of low temperature over-pressure (LTOP) event which challenges the BWR RPV integrity most severely is also assumed as the loading condition for conducting the PFM analyses. Further, the effects of performance of inservice inspection are also studied to determine the benefit of the costly inspection effort. The computed low probability of failure indicates that the analyzed RPVs can provide sufficient reliability even without performing any inservice inspection on the circumferential welds. It also indicates that performing the inservice inspections can not promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on RPV degradation and also be helpful for the regulation of BWR plants in Taiwan.

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