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This work shows the impact of potential displacements of the fuel assembly positions in the reactor core on the signal values of the ex-core instrumentation of a KWU – type pressurized water reactor in order to understand in detail the impact on the calibration factor of excore detectors. This was done with a range of Monte Carlo calculations which simulated the detailed geometrical effect by stepwise changing the positions of fuel assemblies for selected, conservative scenarios. First, criticality calculations were carried out for the chosen core configurations and corresponding surface sources on the core barrel were determined. In these calculations, the distances were varied between the fuel assemblies which were in the line of sight of the ex-core instrumentation. A maximal change of the fluxes on the surface of the core barrel of 4%/mm could be calculated under conservative assumptions for the combination of displaced fuel assemblies. In addition, a dependence of this effect as a function of cycle burnup was analyzed. In a second step transport calculations for the ionization chambers were performed using the surface sources. An increase of the reaction rate at the chambers of up to 3%/mm has been calculated.

This work shows the impact of possible variations of the core geometry on the signal values of the ex-core instrumentation of a pressurized water reactor (PWR). Based on the potential influence on the signals, neutron fluxes have been determined for selected fuel assembly displacements using stationary Monte Carlo calculations. First, critical calculations were carried out for the chosen core configurations and corresponding surface sources on the core barrel were determined. In a second step transport calculations for the ionization chambers were performed using the surface sources. Typical data of a German PWR were used for the investigations. Variation of fuel assembly positions in the outer region of the core were studied, which were directly opposite to the ex-core instrumentation. An increase of the neutron flux at the chambers of up to 4% has been calculated for a change in position of 1 mm between some fuel assemblies. The reason is an improved moderation of neutrons because of the larger water gap. This causes an increase in local power which leads to a greater leakage of neutrons over the core barrel. In addition, a dependence of this effect as a function of cycle burnup was analyzed.

[EN] In nuclear safety field, neutronic and thermalhydraulic codes performance is an important issue. New capabilities implementation, as well as models and tools improvements are a significant part of the community effort in looking for better Nuclear Power Plants (NPP) designs. A procedure to analyze the PWR response to local deviations on neutronic or thermalhydraulic parameters is being developed. This procedure includes the simulation of Incore and Excore neutron flux detectors signals. A control rod drop real plant transient is used to validate the used codes and their new capabilities. Cross-section data are obtained by means of the SIMTAB methodology. Detailed thermalhydraulic models were developed: RELAP5 and TRACE models simulate three different azimuthal zones. Besides, TRACE model is performed with a fully 3D core, thus, the cross-flow can be obtained. A cartesian vessel represents the fuel assemblies and a cylindrical vessel the bypass and downcomer. Simulated detectors signals are obtained and compared with the real data collected during a control rod drop trial at a PWR NPP and also with data obtained with SIMULATE-3K code.

Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The development of these methods is restricted to the compulsory use of fire probabilistic safety assessment (PSA) models. The first method is a fire protection systems and key safety functions unavailability matrix which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. Specific selection and quantification methodologies have been developed to obtain the matrices. The Monte Carlo method has been used to assess the uncertainty of the unavailability matrix. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building., Peer Reviewed, Postprint (published version)