Conventional BWR with Active Safety Systems Simulator

The purpose of this 1300 MWe boiling water reactor simulator is educational - a teaching tool for university professors and engineers involved in teaching various topics related to nuclear power. It was developed by CTI Simulation International Corporation in 2008.

The conventional BWR simulator operational specifics are listed as follows:

• The plant model is a typical 1300 MWe BWR with internal recirculation pumps and fine motion control rod drives.
• There is a number of BWR plants in the world that belong to this category, such as for example: Susquehanna in the USA, Gundremmingen in Germany, and Shika in Japan.
• The simulator can be executed on a personal computer, to operate essentially in real time, and have a dynamic response with sufficient fidelity
• The simulator provides a user–machine interface that mimics the actual control panel instrumentation, including the plant display system, and more importantly, allows user interaction with the simulator during the operation of the plant events.
• The emphasis in developing the simulation models was on giving the desired level of realism to the user in being able to display all plant parameters that are critical to operation, including the ones that characterize the main process, control and protective systems.
• The appropriate parameters and input–output relationships are assigned to each model as demanded by a particular system application.
• Parameter monitoring and plant operator controls are represented in a virtually identical manner on the simulator. Control panel instruments and control devices, such as push–buttons and hand–switches, are shown as stylized pictures, and are operated via special pop–up menus and dialog boxes in response to user inputs.
• There is no possibility to model severe accidents.

Normal Operation

• Power Reduction/Increase
• Normal Reactor Trip

Malfunction Transient Events

• Increasing and Decreasing Core Flow Due to Flow Control Malfunctions
• Inadvertent Withdrawal of One Bank of Control Rods
• Inadvertent Insertion of One Bank of Control Rods
• Inadvertent Reactor Isolation
• Power Loss to 3 Reactor Internal Pumps (RIPs)
• Reactor Bottom Break
• Loss of Both Feedwater Pumps
• Loss of Feedwater Heating
• Reactor Feedwater Level Control Valve Fails Open
• Safety Valves on One Main Steam Line Fail Open
• Steam Line Break Inside Drywell
• Feedwater Line Break Inside Drywel