Siegfried Mittag

Comparative study of a boron dilution scenario in VVER reactors

Abstract Subsequent studies have identified many scenarios, which can lead to reactivity excursions due to boron dilution. The comparative study, presented in this paper, deals with the so-called “restart of the first reactor coolant pump’’ scenario and its reactor-dynamic consequences for both Russian designed VVER reactor types, VVER-440 and VVER-1000. The transient simulations were performed using the three-dimensional core dynamics code DYN3D. The DYN3D modeling features, including recent developments, as well as the cross-section methodology involved in these calculations, are described. The analyzed accident scenario is outlined together with the assumptions made. The results of core response in this boron dilution accident for both VVER reactors are compared within the ranges, determined by the two reactivity values of interest: the criticality limit and the reactivity initiated accident (RIA) limit.

Discontinuity factors for non-multiplying material in two-dimensional hexagonal reactor geometry

Abstract On the basis of methods developed recently for square-fuel-assembly reactor cores, discontinuity factors for hexagonal VVER (Russian PWR) control absorbers and reflector nodes have been derived. Partial currents from heterogeneous multi-group transport calculations are used for the determination of the discontinuity factors. As shown by suitable benchmark calculations, the application of these quantities in the two-group nodal diffusion code DYN3D clearly improves the results of assembly-power predictions. The advantage of reflector diffusion parameters, including discontinuity factors, over conventional albedos has been demonstrated.

Influence of System and Neutron-Kinetic Parameter Variations on an Anticipated Transient Without SCRAM in a PWR

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.Copyright