Notch Ductility, Tensile and Neutron Spectrum Analyses of PM-2A Reactor Pressure Vessel

2009 ◽  
pp. 135-135-21
Author(s):  
C. Z. Serpan ◽  
H. E. Watson
Keyword(s):  
Neutron Spectrum ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

Yankee Reactor Pressure Vessel Surveillance: Notch Ductility Performance of Vessel Steel and Maximum Service Fluence Determined From Exposure During Cores II, III, and IV

1967 ◽  
Vol 89 (4) ◽  
pp. 897-910 ◽  
Author(s):  
C. Z. Serpan ◽  
J. R. Hawthorne
Keyword(s):  
Neutron Spectrum ◽  
Pressure Vessel ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Reactor Pressure Vessel ◽  
Naval Research ◽  
Surveillance Program ◽  
Vessel Steel ◽  
Fuel Cycles ◽  
Reactor Pressure

Charpy V-notch specimens representative of one of the several heats of A302-B steel forming the Yankee reactor pressure vessel, and irradiated as part of the Yankee surveillance program, have been tested by the Naval Research Laboratory. Specimens of this particular heat, irradiated in near-core (accelerated) as well as in vessel wall locations, showed more embrittlement than did specimens of a reference steel heat of the same nominal A 302-B composition irradiated simultaneously in the same surveillance capsules. Those specimens from both the Yankee vessel heat and the reference heat irradiated at the vessel wall location depicted a higher damage rate than that for the accelerated location. The cause of this difference in embrittlement response could not be attributed to an effect of cyclic, service irradiation temperatures, but could be traced to a qualitative relationship of thermal to fast (>1 Mev) neutron fluxes. This ratio was in excess of about 9:1 at the vessel wall location versus a ratio less than about 9:1 for the accelerated location. The computation of a maximum service fluence of 1.46 × 1019 n/cm2 >0.5 Mev was made possible by establishment of the neutron spectrum at the reactor vessel wall using computer calculations. The maximum fluence derived by this technique compared favorably with another value given by an independently-developed calculated neutron spectrum. The NRL computed service fluence in concert with the embrittlement data projects a maximum transition temperature increase of 265 deg F, a level of embrittlement considered acceptable for the Yankee reactor vessel after thirty fuel cycles of operation at 600 MW(t).

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

2014 ◽  
Vol 10 (1) ◽  
pp. 123-127 ◽  
Author(s):  
Gyeong-Geun Lee ◽  
Yong-Bok Lee ◽  
Min-Chul Kim ◽  
Junhyun Kwon
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Rate Theory ◽  
Reactor Pressure

The effects of thermal stress on the crack propagation in AP1000 reactor pressure vessel

2020 ◽  
Vol 110 ◽  
pp. 102798
Author(s):  
KaiTai Liu ◽  
Mei Huang ◽  
JunJie Lin ◽  
HaiPeng Jiang ◽  
BoXue Wang ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Crack Propagation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Ap1000 Reactor ◽  
Reactor Pressure

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.

The ageing behavior and the correlation between hardness and coercivity of Cu-rich reactor pressure vessel model steels

2021 ◽  
Vol 527 ◽  
pp. 167698
Author(s):  
Xuejiao Wang ◽  
Wenjiang Qiang ◽  
Guogang Shu ◽  
Junwei Qiao ◽  
Yucheng Wu
Keyword(s):  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

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.

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