Conventional Two Loop PWR Simulator (PCTRAN)

PCTRAN is a two loop PWR reactor transient and accident simulator. Since its first release in 1985, Micro–Simulation Technology has been constantly upgrading its performance and expanding its capabilities. The PCTRAN simulator operational specifics are listed as follows.:

  • Graphic User Interface (GUI) adheres strictly to the specifications of the Microsoft Windows environment. Data input/output are in MS Office’s Access database format
  • The plant model is a generic two loop PWR with inverted U–bend steam generators and dry containment system, such as the Westinghouse, Framatome or KWU designs with a thermal output in the range of 1800 MWt (600 MWe). Examples of these types of reactors include: Point Beach, Kewaunee, Prairie Island and Ginna in the United States, Mihama1 in Japan, Krsko in Slovenia, Angra1 in Brazil, and ChinShan2 in China, as found in the IAEA PRIS Database.
  • TRIGA model is available for demonstrating the concepts of neutron multiplication, rod control to criticality, feedback, decay heat, and effects of poisoning.

The PCTRAN simulator can address (to a certain extent) the behavior of the plan under severe accidents, as follows:

  • TMI–2 accident: this accident is simulated by triggering combination of loss of condensate pump, main feedwater pumps and disabling the auxiliary feedwater in thus allowing the users to analyze: steam generator level, peak cladding temperature, system pressure changing over time during the accident.
  • Station Blackout (SBO), both off–site AC and on–site diesel power are lost: only DC operated pressurizer and steam generator Pilot Operated Relief Valve (PORV) are operating to relieve pressure. Prolonged SBO may lead to vessel failure as well as core melt.
  • Large Break without emergency core cooling system (ECCS): 2000 cm2 cold leg severe accident is modeled by disabling the accumulators, HPI and LPI pumps. The core is rapidly exposed and starts to melt, then collapses and melts through the vessel bottom. Users may observe corium concrete interaction and aerosol generation in the containment.

Normal Operation

  • Power Reduction/Increase
  • Normal Reactor Trip

Malfunction Transient Events

  • Power Reduction/Increase
  • Normal Reactor Trip
  • Uncontrolled Rod Bank Withdrawal
  • Hot Full Power Rod Drop
  • Moderator Dilution
  • Startup of an Inactive RCP
  • Reduction in Feedwater Enthalpy
  • Excessive Load Increase
  • Loss of Reactor Coolant Flow
  • Turbine Trip
  • Loss of Normal Feedwater
  • Steam Generator Tube Rupture

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