Thermal-Hydraulic Analysis Approach

COBRAIIIc is a widely used code for the thermal-hydraulic analysis of nuclear reactor cores. Its approach is based on subchannel analyses where conservation equations are axially solved and coupled through mixing coefficients (Todreas Kazimi 2001). 

The core design procedure consists of two parts, nuclear design and thermal-hydraulic analysis. The former is based on the fine-mesh multi-group neutron diffusion solution. The latter is based on single channel analyses for the average and hot channels of all the fuel assemblies. This approach is the same as that in the Super LWR design. 

This report presents our approach for calculation of the Task. Our numerical code THAMES is the three-dimensional finite element simulator of fully coupled processes. First, we defined the input data for THAMES from the supplied properties of FEBEX bentonite. After calibrations of some parameters such as thermal vapour diffusivity, the analysis that treats fully coupled thermal, hydraulic and mechanical processes was carried out. 

For some aspects of model uncertainties, well-known quantification methods are available. A Bayesian approach might be practicable, for instance, to quantify the uncertainty on the probability model to apply for the stochastic failure behaviour of system components. Monte Carlo analysis might be appropriate to quantify the uncertainty resulting from the application of thermal-hydraulics codes. 

The analysis should use a logical approach which models how the event sequences progress from core damage to a radiological release. This is usually done by event tree analysis which models the accident sequence in a number of time frames and uses a set of nodal questions to model the sequence of events which occur. The construction of the event trees needs to be supported by thermal-hydraulic calculations and modelling of fission product release and transport inside the containment. 

Best-estimate thermal hydraulics and neutronic codes are used to perform the analysis for the LOCA and other DBAs. It is required to identify and assess the associated uncertainties in the analytical models and input data in such a way that the uncertainties in the calculations can be quantified when the results are compared with the acceptance criteria, giving adequate assurance that the acceptance criteria are met. Mature best-estimate codes are widely available aroimd the world, an extensive database exists for nearly all power reactor designs, and best-estimate plant calculations are well documented. The analysis of accidents with best-estimate codes using combinations of best-estimate and conservative inputs is particularly appropriate since this approach provides some estimates of the uncertainties in the overall plant behavior. These estimates can then be compared with the uncertainty estimates developed through relevant activities in different countries in code validation, as well as studies on representation and uncertainties in plant data, to help establish confidence in the predicted behavior of the plant. This approach is dependent on the continued emphasis on activities in development and validation of best-estimate codes to ensure that such codes can be used with a high degree of confidence. 

The development of methods of calculation for thermal-hydraulic design has proceeded in close association with the work on nuclear design methods. The complete system of digital codes now available for steady state performance analysis is known as PATRIARCH (see Pig. 1). Descriptions of the functions of the reactor physics codes appear in a companion paper. The approaches used in the thermal-hydraulic codes will be reviewed briefly below. In the space available, it is only possible to outline the basis of the more important codes. 

Although the coolant flow in the Super LWR is single-phase, the coolant enthalpy and therefore the density change substantially in the core because the coolant flow rate per thermal power in the Super LWR core is less than one eighth of LWR cores. Thus, the Super LWR can be susceptible to flow oscillations as the BWRs are. In Sect. 5.4, thermal hydraulic stability of the Super LWR is analyzed with the frequency domain approach. The analysis includes both supercritical and subcritical pressure conditions. 

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