The complete failure of the reactor scram system upon request during an operational transient is called anticipated transient without scram (ATWS). According to the German regulatory guidelines, postulated ATWS events have to be analyzed with regard to their consequences on the safety of nuclear power plants. Since the course of ATWS transients is determined by a strong interaction of the neutron kinetics with the thermal hydraulics of the system, coupled 3D neutron kinetic/thermal hydraulic code systems are adequate tools for the analysis of such transients. The coupled code system DYN3D/ATHLET developed at FZD is applied to the analysis of an ATWS transient. The objective of the present work is to perform a best-estimate analysis with consequent use of a 3D neutron kinetic code (DYN3D) in combination with an advanced thermal hydraulic system code (ATHLET) together with a quantification of differences in the course and the results of transients, which arise from the uncertainties of thermal hydraulic and neutron-physical conditions. Typically, the complete failure of the main feed water supply is assumed to be the bounding ATWS event with regard to the maximum primary coolant pressure, which can be reached during the transient. The limitation of the coolant pressure is a precondition for the integrity of the primary circuit. The situation is aggravated if the main coolant pumps remain in operation. For this particular transient, the influence of different thermal hydraulic and neutron-physical conditions on the course of the transient was analyzed. In a number of code runs all systems which have an influence on the course of the transient were varied. These are the auxiliary boration system and the auxiliary feed water supply. Further, the influence of the modeling of the pressurizer safety and relief valves as well as the steam bypass system on the secondary side was assessed. The variation of the pressurizer relief and safety valve behavior has the biggest influence on the primary circuit coolant pressure. In the second part, two different core loading patterns were generated for the analyses by varying the number of MOX (mixed oxide) fuel assemblies (FA) in the core. The basic core loading contains 64 MOX FA. All these MOX FA were replaced by standard uranium oxide FA. The presence of MOX in the core has a remarkable influence on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behavior of the coolant pressure in the first part of the transient. It has been demonstrated that the pressure maximum decreases with growing MOX portion in the core. The maximum pressure determined in the calculations with variation of system and neutron-physical boundary conditions does not reach the allowed limit for the primary circuit.

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