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Extra resources for Accid. Anal. - Nucl. Powerplants w. Graphite-Moderated BW RBMK Reactors
Thus, an energy criterion that is only slightly sensitive to the heating rate may be adopted for the thermomechanical code employed for calculating the deformation and for assessing pressure tube integrity, as shown in Fig. 7. This criterion is a specific rupture strain power ji (W/kg) which is determined by the stress intensity si, material density rw and strain rate intensity zi: ji = s iz i rw The criterion curve in Fig. 7 is obtained by approximation of empirical data by two conjugated sixth degree polynomials: k =6 j ip = Âa kT k k =0 with polynomial coefficients ak given as in Table 6.
The choice is made within the range of possible values depending on the operating mode and on the ranges of the parameters in question. Such ranges may include technological tolerances, calculation and/or measurement errors. Depending on the purpose of a specific analysis, the input parameters can be set at different ends of the range. For instance, for a conservative assessment of the coolability of the core channels during a LOCA, the input parameters will be set to have the smallest coolant inventory in the circulation circuit on the one hand, and the greatest rate of discharge through the break on the other, the longest possible delays in ECCS water delivery and its lowest possible flow rate.
For this reason, it is appropriate to use a 3-D neutronics code with a built-in multichannel thermohydraulic model. In addition, a 3-D neutronics code should be employed for compiling input data for point neutron kinetics (commonly involved in thermohydraulic codes) when events such as LOCAs are analysed. Considering the fact that the MCC is divided into two symmetric loops, a pipe break in one of these directly affects the thermohydraulic and neutronic behaviour on one side of the core. The resulting asymmetry in reactivity and power has to be assessed by a 3-D code.
Accid. Anal. - Nucl. Powerplants w. Graphite-Moderated BW RBMK Reactors