Wadim Jaeger

Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by Means of Steady State BFBT Data

The subject of the present paper is the uncertainty and sensitivity studies for steady state BFBT results including pressure drop and void fraction measurements. The investigations are performed with TRACE (version 5.0 patch 2), thermal hydraulic modeling, and SUSA and DAKOTA; both tools are for the evaluation of uncertainties and sensitivities. For this purpose, the NUPEC BFBT experimental data base is used. The advantage of applying two different uncertainty and sensitivity tools in combination with TRACE is that the user effect can be excluded. Since in both cases the TRACE model of the BFBT bundle is identical the differences in the results are related to the capabilities of the uncertainty and sensitivity tools. The reference results with TRACE show that the code is very well able to represent both single- and two-phase flows even though it is a 1D coarse mesh system code. For selected cases, an uncertainty study was performed. Even though a reduced number of uncertain parameters are considered in the DAKOTA investigation, compared to the one with SUSA, similar results are obtained. The results indicate also that even small parameter variations can yield to rather large variations of the selected output parameters.

On the Uncertainty and Sensitivity Analysis of Experiments With Supercritical Water With TRACE and SUSA

This study was performed at the Institute for Neutron Physics and Reactor Technology at Karlsruhe Institute of Technology and is addressed to establish the combined usage of the best-estimate code TRACE and the uncertainty and sensitivity analysis tool SUSA, for safety related investigations of current and future nuclear energy systems. In the frame of this paper the applicability to supercritical water related investigations is covered. Several Nusselt correlation for the heat transfer to supercritical water are available and have been evaluated in previous investigations but not one of them gave satisfying results. Hence, the consideration of uncertainty and sensitivity measures applied to the topic of heat transfer seems to be an appropriate way. In a first step a post-test analysis of an experiment was conducted. Results showed that with the help of uncertainty and sensitivity methods parameters which affect the results most could be identified. The most important parameter was of course the Nusselt correlation. In addition to the identification of important parameters, the experimental results were enveloped by the calculated results. That means, in the sense of safety related evaluation of designs for reactors operated with supercritical water, that key parameters (cladding temperature) can be calculated with a certain confidence.Copyright

Validation of TRACE in the Field of Liquid Metal Heat Transfer

The validation of system codes like TRACE is an ongoing task especially in areas with limited or almost no application like liquid metal flow. Therefore, extensive validation efforts are necessary to increase the confidence in the code predictions. TRACE has been successfully validated and applied to lead-alloy cooled systems. The results gained with lead-alloy coolants could be extrapolated to other liquid metals with the necessary care. Nevertheless, dedicated investigations with the different liquid metals are mandatory to confirm the extrapolations. In the present case, the validation work focuses on liquid metal heat transfer in pipes and rod bundles under forced convection. To take advantage of a greater data base, several liquid metals have been implemented into the code. In addition, new coolants allow supporting analysis of liquid metals loops which are in the design or construction stage. Concerning the validation, several experiments have been found, conducted by other investigators, which are modeled with the modified TRACE version. The results indicate that the chosen heat transfer models for pipe and bundle flow are applicable. In case of deviations, physical sound reasons can be provided to explain them.Copyright

Uncertainty and Sensitivity Study in the Frame of TRACE Validation for Reflood Experiment

Abstract The validation of computer codes related to the thermal-hydraulic analyses of nuclear reactors is a challenging undertaking because of the complexity of the phenomena involved, e.g., boiling, condensation, and mixing. In the frame of the ongoing validation of the best-estimate system code TRACE, the present paper focuses on the phenomena taking place during the quenching of the hot surface of the fuel rod simulator with cold water. Since TRACE describes the physical phenomena with empirical correlations derived from experiments, it is necessary to ensure that these correlations are valid if applied to similar experiments but different boundary conditions. By means of an uncertainty and sensitivity study, the influence of the empirical models and their associated uncertainties on selected output parameters is quantified and the parameters with the largest sensitivity are evaluated.

Thermal-Hydraulic Evaluation of an LBE Cooled 19-Pin Bundle in the Frame of TRACE Validation

In the frame of the validation and verification of the thermal hydraulic system code TRACE for lead alloy cooled systems, a 19 pin bundle is analyzed, in particular, the influence of the spacers. TRACE is updated and improved in order to be used for the analyses of liquid metal cooled reactors. For the present investigation, isothermal hydraulic experiments, carried out at the Karlsruhe Lead Laboratory, are used to validate the pressure losses related to wall friction and spacers. In this experimental set-up, a 19 rod bundle, cooled by liquid lead-bismuth eutectics (LBE) is used. In addition, thermal analyses are performed in order to evaluate the axial temperature profile of coolant and cladding and to validate the chosen heat transfer models. Since the thermal experiments have not been performed yet, the validation is done by comparing TRACE results to CFD results following a unique approach for the turbulent Prandtl number. Special emphasize is paid to the validation of the models to account for heat transfer enhancement at and in the vicinity of spacers due to increased turbulences.Copyright

Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

The Institute of Neutron Physics and Reactor Technology (INR) is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV) as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

Investigation of Local Heat Transfer Enhancement in Generic Liquid Metal–Cooled Fuel Assemblies with Empirical Models [in press]

Abstract The heat transfer in liquid metal–cooled rod bundles is modeled with a knowledge-based best-estimate system code. Thereby, the focus is on the heat transfer enhancement due to flow perturbations. These perturbations are caused by local geometrical variations, such as sudden expansions and contractions, in the flow channel. The accurate calculation of the heat transfer is important for the safety demonstration of, e.g., subassemblies. Safety-related parameters, such as fluid and wall temperature, have to satisfy certain limits during normal and off-normal operation as well as during accidents. Up to now, fully developed flow is assumed for heat transfer in liquid metal–cooled rod bundles. The effects of local heat transfer enhancements were ignored in best-estimate system codes. The currently used empirical heat transfer models are functions of the Péclet number only. Several experimental and numerical investigations show that flow perturbations induce higher heat transfer due to increased turbulences, accelerated flows, and secondary motions. In this paper, the effects of the entrance region and the presence of spacer grids on the heat transfer are investigated. Empirical models for that are selected and applied. These empirical models are functions of the Péclet number, the geometrical perturbation, and the distance from the perturbation in the flowing direction. The calculated heat transfer coefficients at the bundle entrance and in the vicinity of spacer grids are twice as high compared to bare rod bundles under a fully developed flow condition without any flow perturbation. Because of the higher heat transfer, lower wall temperatures are to be expected. This provides additional safety margins during normal and off-normal operation as well as during accidents. Furthermore, the considerable increase of heat transfer shows that existing perturbations have to be considered to obtain accurate and reliable results.

On the Applicability of Pool Boiling Models in TRACE for the Evaluation of a Counter-Current LBE-Water Heat Exchanger

A counter current LBE-Water heat exchanger is investigated with the system code TRACE with emphasize on the water side. Due to the liquid metal coolant on the outer side of the heat exchanger the present boundary conditions for the water side (pressure, mass flux, heat flux, etc.) might be out of range of normal LWR application and need therefore additional validation. The review of the physical models of TRACE, mainly the ones for pool boiling, indicated that these models might not be applicable to the present heat exchanger design. Alternative pool boiling models have been implemented into TRACE, namely the ones of Forster-Zuber and of Labuntsov. The comparison shows that the heat transfer coefficient of the original model is about 30% higher than with the alternative models but with a wall heat flux which is almost identical for the three models. This is caused by the much lower heat transfer coefficient on the LBE side which dictates the global heat transfer. The resulting wall temperature and the power which can be transferred from the hot LBE side to the water are lower for the alternative models. That indicates that with the standard model too optimistic results might be calculated.Copyright