The Euler method is used for integration and first order lag, and the Runge-Kutta method is used for the steamline dynamics.
Major assumptions and simplifications are summarised in the following;
(1) Pressure is assumed to be uniilmn in the reactor vessel. Differential pressures can be calculated optionally in two regions, the lower part of the free surface and the rest of the vesseL
(2) In the thermal-hydraulic model for reactor vessel, saturated coolant region (from core to feedwater sparger) are divided into five regions, core, upper plenum, separator, steam dome, free surface region from feedwater sparger to free surface).
(3) A bubble separation model is used in the subcooled coolant region downcomer, recirculation pipe and lower plenum) when coolant becomes saturated.
(4) The core is divided into stacked several nodes to take into consideration the void distribution in the axial direction.
(5) Empirical correlations on slip ratio for normal coolant flow and slip velocity for slow coolant velocity after recirculation pump trip are used to simulate the void behavior.
(6) Coolant mass change in the separator is calculated from water level, inlet steam and total flows by the correlation obtained from the experimental data.
(7) Recirculation loops can be simulated with or without jet pumps.
(8) Core inlet flow and recirculation flow are calculated on the basis of momentum equation obtained by the analogy with electrical circuits. Other flows in the saturated regions are calculated from mass and energy equations.
(9) First order lag representations are used in order to apply static correlations to dynamic behaviors.
Other characteristics of the code are:
(1) A point kinetic model is used for neutron flux calculation.
(2) In the fuel temperature calculation, the fuel rod is divided into several nodes in the radial and axial direction, gap heat capacity is ignored and the clad is treated as one region in radial direction.
(3) The temporal momentum change of steam flow in main steamline Can be optionally considered.
(4) The control systems fix pressure, feedwater and recirculation flow are simulated.
(5) The thermal-hydraulics in hot channel including MCHFR (Maximum Critical Heat Flux Ratio) can be calculated separately from the calculation of those in whole core.
(6) Emergency core cooling system (ECCS) flow can be calculated.
Typical transients which can be analyzed by BWRDYN are as follows:
(1) Recirculation pump trip
(2) Feedwater pump trip and restart
II) Valve position change or failure
(1) Turbine main stop valve
(2) Turbine control valve
(3) Turbine bypass valve
(4) Relief valve
(5) Safety valve
(6) Main steam isolation valve
(7) Feedwater control valve
(8) Drain level. control valve of BOP model
III) Control system characteristics
(1) Pressure set point change of pressure regulator
(2) Water level set point change of feedwater control system
(3) Flow set point change of recirculation flow control system
(4) Power set point change of recirculation flow control system
(1) Feedwater enthalpy change
(2) Reactivity insertion
(3) Feedwater line isolation of BOP model