Protection against intermediate sodium system overpressure from the sodium/water reaction associated with large leaks within the CRBRP Steam Generators is provided by the sodium/water reaction pressure relief system (SWRPRS). This system consists of rupture disks connected to the intermediate sodium piping adjacent to the inlet to the superheater and outlet from the evaporator modules. The rupture discs relieve into piping that leads to reaction produce separator tanks, which in turn are vented to a centrifugal separator and flare stack arranged to burn hydrogen gas exhausting into the atmosphere. Analyses have been conducted using the TRANSWRAP Computer Code to predict the system pressures and flow rates during the large leak event. Experi...
Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe re...
International audienceSteam Generator (SG) is one of the vital components of sodium cooled fast reac...
Two tests related to a new safety system for a pressurized water reactor were performed with the ROS...
Diffusion membrane type hydrogen detectors are provided for monitoring the sodium exiting each evapo...
Tests will be conducted on the CRBRP prototype steam generator at SCTI to determine the effects of s...
When pressurized water or vapor leaks from a failed heat transfer tube in a steam generator of sodiu...
AbstractThe prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Kor...
A new computational methodology of sodium-water reaction (SWR), which occurs in a steam gen-erator o...
Sensitivity studies on Sodium-Water Reaction (SWR), which will take place in a steam generator of Li...
Due to the Three Mile Island accident in 1979 the Swedish government took the decision in 1986 to im...
In the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experi...
Abstract – A numerical investigation of sodium-water reaction (SWR) phenomenon under a tube bundle c...
During a hypothesized severe accident, a containment building is designed to act as a final barrier ...
The MECTUB code was developed to evaluate the risk of swelling and bursting of Steam Generator (SG) ...
The objective of the small-scale group of tests is to demonstrate that sodium will drain from the su...
Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe re...
International audienceSteam Generator (SG) is one of the vital components of sodium cooled fast reac...
Two tests related to a new safety system for a pressurized water reactor were performed with the ROS...
Diffusion membrane type hydrogen detectors are provided for monitoring the sodium exiting each evapo...
Tests will be conducted on the CRBRP prototype steam generator at SCTI to determine the effects of s...
When pressurized water or vapor leaks from a failed heat transfer tube in a steam generator of sodiu...
AbstractThe prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Kor...
A new computational methodology of sodium-water reaction (SWR), which occurs in a steam gen-erator o...
Sensitivity studies on Sodium-Water Reaction (SWR), which will take place in a steam generator of Li...
Due to the Three Mile Island accident in 1979 the Swedish government took the decision in 1986 to im...
In the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experi...
Abstract – A numerical investigation of sodium-water reaction (SWR) phenomenon under a tube bundle c...
During a hypothesized severe accident, a containment building is designed to act as a final barrier ...
The MECTUB code was developed to evaluate the risk of swelling and bursting of Steam Generator (SG) ...
The objective of the small-scale group of tests is to demonstrate that sodium will drain from the su...
Natural circulation flows can develop within a reactor coolant system (RCS) during certain severe re...
International audienceSteam Generator (SG) is one of the vital components of sodium cooled fast reac...
Two tests related to a new safety system for a pressurized water reactor were performed with the ROS...