Core thermal hydraulics

Central UO2 temperature The maximum UO2 temperature will occur in new fuel operating at the maximum linear heat generation rate of 13.4 kW/ft (44 kW/m). Based on published conductivity data, the maximum temperature is approximately 3400°F (1871°C). 

Core orificing Fixed orifices accomplish control of core flow distribution among the fuel assemblies. These orifices are located in the fuel support pieces and are not affected by fuel assembly removal and replacement. 

The core is divided into two orifice zones. The outer zone of fuel assemblies, located near the core periphery, has more restrictive orifices than the inner zone, so flow to the higher power fuel assemblies is increased. The orificing of all fuel assemblies increases the flow stability margin. 

Comparisons of fhe analytical models used wifh fuel assembly design details such as fuel-rod-to-fuel-rod and fuel-rod-to-fuel-assembly-channel clearances and spacer configurations have been made to ensure that the computer programs adequately represent the actual core and fuel design, and fhat design correlafions are applicable. 

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March-Leuba (1990) presented radial nodalization effects on the stability calculations. March-Leuba and Blakeman (1991) reported on out-of-phase power instabilities in BWRs. BWR stability analyses were reported by Anegawa et al. (1990) and by Haga et al. (1990). The experience and safety significance of BWR core-thermal-hydraulic stability was presented by Pfefferlen et al. (1990). 

Pfefferlen, H., R. Raush, and G. Watford, 1990, BWR Core Thermal-Hydraulic Stability Experience and Safety Significance, Proc. Int. Workshop on BWR Instability. Hottsville, NY, CSNI Rep. 178, OECD-NEA, Paris. (6)... 

Thermal hydraulic design. The preliminary core thermal hydraulic design reported here needs to be continued using multidimensional methods to establish the heat removal capability and coolant pump sizing for the target power level and for assessing the reactor s performance in anticipated transients. 

Core thermal hydraulic values are summarized in Table XII. The coolant inlet temperature and temperature rise are 800° and 300°F, respectively, giving a bulk outlet temperature of 1100°F, which is considered... 

The computer codes which are used to carry out the anticipated operational occurrences and DBA analysis should be properly verified and validated. This includes the codes used to predict the behaviour of the reactor core, thermal-hydraulic codes and the radiological release and consequence codes. In addition, the analysts and users of the codes should be suitably qualified, experienced and trained. 

Open-lattice core thermal hydraulics. The velocities through the core from natural circulation are low (e.g. u mean 0.9 m/s). The core design has a significant power spatial distribution (peaking factor of 1.63) that gives rise to significant temperature and velocity profiles across the core ... 

MELCOR has been used to model experiment LP-FP-2 [25, 26, 27, 28, 29, 30, 31], which simulates many of the primary system and core thermal/hydraulic conditions that would be expected during a PWR V-sequence. The relatively large scale of the test and the extensive instrumentation used make the LP-FP-2 experiment an important integral source of data for qualifying severe accident code predictive capabilities. 

From Table 2, mass flow distributions to the three core Mets exhibit (see Figure 2) a nearly inverted profile to those specified for the in-core thermal/hydraulic design. That is, more flow is going to the innermost fuel region of the core than was assumed by the in-core analysis), leaving the outer region with less flow. 

The orifices in BWRs are mainly used for improving the core thermal-hydraulic stabilities. Generally, the BWR channel stability improves when the pressure drops and inertia in the single-phase flow region are increased. This is why inlet orifices are used in BWRs. For LMFBRs, inlet orifices are used to control the coolant flow rate to the fuel assemblies to effectively cool the fuel. 

The single channel analysis model for the core thermal-hydraulic calculations does not consider the pressure drops and only takes into account the conservations of... 

S Applying the Single Channel Model to Core Thermal-Hydraulic Calculations... 

The core thermal-hydraulic calculations are based on the single channel analysis model. On the other hand, the three-dimensional core power distribution is obtained by COREBN for the calculation mesh described in Fig. 2.20 [9]. In the core thermal-hydraulic calculations, the neutron flux calculation mesh of the COREBN is assumed to compose a fuel channel group. The fuel channels in this fuel channel group are assumed to be identical. 

Figure 2.29 [9] shows the core thermal-hydraulic calculations by the single channel model. Each fuel assembly is assumed to be composed of 36 fuel channel... 

Fig. 2.29 Core thermal-hydraulic calculations by the single channel model. (Taken from doctoral thesis of A. Yamaji, the University of Tokyo (2005) [9])...

This section describes the basic design concepts of the Super LWR core including the fuel rod and fuel assembly designs. The core thermal-hydraulic characteristics are unique and strongly coupled with the neutronic characteristics of the core. 

The neutronic feedback involves the neutron kinetics, the fuel dynamics, the core thermal-hydraulics, and the reactivity feedback dynamics. The neutron kinetics affects and is affected by the power generation in the fuel, and is directly responsible for the power perturbations. The fuel dynamics affects and is affected by the fuel surface heat flux, and is responsible for the time delays between power production and the response of coolant flow heating. The core thermal-hydraulics affects the power production and the response of the water density perturbations to fuel surface heat flux perturbations. Finally, the reactivity feedback dynamics is responsible for the feedback reactivity due to water density perturbations and fuel temperature perturbations, and is affected by neutron kinetics. 

The core water level is calculated from a detailed core thermal-hydraulic calculation. The quench front is calculated by a theoretical correlation proposed by Yamanouchi, et al. [24]. The heat transfer coefficient sharply changes by about 2 orders of magnitude in the vicinity of the quench front. In order to prevent the numerical instability caused by the abrupt change in the heat transfer coefficient, the neighboring nodes of the quench front are more finely divided into a size 1/100 of the thickness of the normal node as shown in Fig. 6.21. The flow regimes assumed in the reflood analysis are described in Fig. 6.22. Various heat transfer correlations are prepared according to the flow conditions. Table 6.9 [21] summarizes them. 

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