Analytical approach to estimate the thermal stress distribution of reactor pressure vessel nozzle corners with a constant cooldown rate

2022 ◽  
pp. 104608
Author(s):  
Changsik Oh ◽  
Sangmin Lee ◽  
Myung Jo Jhung ◽  
Nam-Su Huh
Keyword(s):  
Thermal Stress ◽  
Stress Distribution ◽  
Pressure Vessel ◽  
Analytical Approach ◽  
Reactor Pressure Vessel ◽  
Thermal Stress Distribution ◽  
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

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions such as loss-of-coolant accident (LOCA) is a major safety concern. Besides conventional deterministic calculations to justify as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

Thermal-stress modeling of AP1000 reactor pressure vessel due to transient pressurized thermal shock

2017 ◽  
Vol 47 (7) ◽  
pp. 685-691
Author(s):  
ZiQi MA ◽  
HongLiang ZHANG ◽  
PeiFeng HE ◽  
Bin XU ◽  
Ying LUO ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Stress Modeling ◽  
Pressurized Thermal Shock ◽  
Ap1000 Reactor ◽  
Reactor Pressure

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

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
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