The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

Information Fusion Technique for Evaluating Radiation Embrittlement of Reactor Pressure Vessel Steels

2005 ◽  
Vol 127 (1) ◽  
pp. 98-104 ◽  
Author(s):  
J. A. Wang ◽  
S. Konduri ◽  
N. S. V. Rao
Keyword(s):  
Information Fusion ◽  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Time ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Fusion Technique ◽  
Information Fusion Technique ◽  
Reactor Pressure

A new approach that utilizes the information fusion technique was developed to predict the radiation embrittlement of reactor pressure vessel (RPV) steels. The Charpy transition temperature-shift data is used as the primary index of the RPV radiation embrittlement in this study. Six parameters, Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved about 66% and 53% reductions, respectively, in the uncertainties for the update General Electric (GE) Boiling Water Reactor (BWR) plate and weld data compared to the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2 (RG1.99/R2). The implications of irradiation temperature effects for the development of radiation embrittlement models are also discussed.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

Dosimetry Assessments for the Reactor Pressure Vessel and Core Barrel in UK Pressurized Water Reactor (PWR) Plant

2012 ◽  
Vol 9 (4) ◽  
pp. 104016 ◽  
Author(s):  
D. A. Thornton ◽  
D. A. Allen ◽  
A. P. Huggon ◽  
D. J. Picton ◽  
A. T. Robinson ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

Radiation Effects in Boiling Water Reactor Pressure Vessel Steels

1980 ◽  
Vol 102 (2) ◽  
pp. 177-186
Author(s):  
J. N. Kass ◽  
A. J. Giannuzzi ◽  
D. A. Hughes
Keyword(s):  
Pressure Vessel ◽  
Radiation Effects ◽  
Statistical Evaluation ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Shelf Energy ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Effect of neutron irradiation on notch toughness properties of Boiling Water Reactor pressure vessel steels was determined. Samples from several heats of plate, weld metal, and forgings were irradiated to three different fluence levels and tested. A statistical evaluation of the data was conducted to determine regression analysis mean decreases in upper shelf energy and increases in transition temperature versus fluence.

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