Startup Thermal Analysis Code

The plant analysis code developed at the University of Tokyo for the Super LWR with downward flow type water rods is used here. The thermal hydraulic analyses are carried out by using the one-dimensional single-channel model of the fuel 

The thermal calculations are carried out from the core inlet to the core outlet. The inlet coolant temperature and mass flow rate are used as boundary conditions. The temperatures of the coolant in fuel channels and those of the moderator water in the water rods are calculated from the mass and energy conservation equations. The axial power is assumed to follow a cosine distribution. The radial power distribution in the fuel assembly is not considered. The steady-state temperature distributions are assumed in the fuel pellet, fuel cladding, and the gap. The thermal power generated in the reactor is to be consumed among the turbines, the condenser, and the feedwater heaters. The calculations are carried out iteratively until the solutions are convergent to steady-state values. 

From the coolant and moderator temperature distributions, the fuel and cladding temperatures are calculated using one-dimensional heat transfer equations. The heat transfer between the fuel channel and the water rod and the heat transfer between the fuel pellet and the coolant are considered. 

For supercritical pressures, the heat transfer coefflcients are calculated by using the Oka-Koshizuka correlation (see Sect. 2.3.2.2). 

Mode Heat transfer type Heat transfer correlation 

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The plant dynamics code for the analysis of plant control and startup thermal considerations are described in ref. [115]. The subchannel analysis code and the analysis are found in refs. [116, 117]. Thermal-hydraulic and coupled stability calculations at supercritical and at subcritical pressure as well as startup considerations are described in ref. [118]. 

It was revealed that most of the events due to a work planning problem where a work procedure is provided occurred during low-power states or startup operations. The reason for this can be inferred as the variable characteristics of plant configuration and dynamics of the low-power states of NPP, which may cause the identification of human error potentials and prediction of physical transition to be difficult. Therefore, the identification of human error possibilities or potentials during low-power states seems not to be an easy task to be accomplished by a list of simple checklist items, but belongs to a hard task that requires a careful investigation on the potential of human error by a concerted effort between experts in plant systems and hiunan errors, and, as necessary, may requires the use of thermal-hydraulic and reactor analysis computer codes. 

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