A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period in...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
[[abstract]]Reports on a series of small-break LOCA analyses for a typical Westinghouse 3-loop PWR, ...
A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has b...
Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In...
Debris generated during the blowdown phase of a Loss of Coolant Accident (LOCA) is a big concern in ...
The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safe...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
This report revision incorporates new experimental evidence regarding AP600 behavior during small br...
Abstract During the operation lifetime of a Nuclear Research Reactors, the frequency of occurrence...
An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel ...
The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems ...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
[[abstract]]Reports on a series of small-break LOCA analyses for a typical Westinghouse 3-loop PWR, ...
A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has b...
Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In...
Debris generated during the blowdown phase of a Loss of Coolant Accident (LOCA) is a big concern in ...
The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safe...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a ...
This report revision incorporates new experimental evidence regarding AP600 behavior during small br...
Abstract During the operation lifetime of a Nuclear Research Reactors, the frequency of occurrence...
An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel ...
The Westinghouse AP600 reactor is an advanced pressurized water reactor with passive safety systems ...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
Thermal-hydraulic analysis is a key part in support of regulatory work and nuclear power plant desig...
[[abstract]]Reports on a series of small-break LOCA analyses for a typical Westinghouse 3-loop PWR, ...