Abstract
A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k{sub I}) with the crack initiation toughness (k{sub Ic}) and crack arrest toughness (k{sub Ic}), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author).
Hirano, Masashi;
[1]
Kohsaka, Atsuo
- Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
Citation Formats
Hirano, Masashi, and Kohsaka, Atsuo.
Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events.
Japan: N. p.,
1992.
Web.
Hirano, Masashi, & Kohsaka, Atsuo.
Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events.
Japan.
Hirano, Masashi, and Kohsaka, Atsuo.
1992.
"Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events."
Japan.
@misc{etde_10149592,
title = {Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events}
author = {Hirano, Masashi, and Kohsaka, Atsuo}
abstractNote = {A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k{sub I}) with the crack initiation toughness (k{sub Ic}) and crack arrest toughness (k{sub Ic}), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author).}
place = {Japan}
year = {1992}
month = {Feb}
}
title = {Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events}
author = {Hirano, Masashi, and Kohsaka, Atsuo}
abstractNote = {A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k{sub I}) with the crack initiation toughness (k{sub Ic}) and crack arrest toughness (k{sub Ic}), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author).}
place = {Japan}
year = {1992}
month = {Feb}
}