Victor Sanchez

Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT). The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

SCALE Modeling of Selected Neutronics Test Problems within the OECD UAM LWR’s Benchmark

The OECD UAM Benchmark was launched in 2005 with the objective of determining the uncertainty in the simulation of Light Water Reactors (LWRs) system calculations at all the stages of the coupled reactor physics—thermal hydraulics modeling. Within the framework of the “Neutronics Phase” of the Benchmark the solutions of some selected test cases at the cell physics and lattice physics levels are presented. The SCALE 6.1 code package has been used for the neutronics modeling of the selected exercises. Sensitivity and Uncertainty analysis (S/U) based on the generalized perturbation theory has been performed in order to assess the uncertainty of the computation of some selected reactor integral parameters due to the uncertainty in the basic nuclear data. As a general trend, it has been found that the main sources of uncertainty are the 238U (n,) and the 239Pu nubar for the UOX- and the MOX-fuelled test cases, respectively. Moreover, the reference solutions for the test cases obtained using Monte Carlo methodologies together with a comparison between deterministic and stochastic solutions are presented.

Parametric and Uncertainty Study of a Counter Current LBE-Water Heat Exchanger With TRACE and SUSA

Due to the variety of existing physical models for the heat transfer and for the description of thermo physical properties, the modeling results of different users for the same design can be different. These discrepancies can be rather big and have therefore a big impact on the thermo hydraulic performance of the investigated design proposals, in the present case a LBE-water counter current heat exchanger. A parametric, and a subsequent uncertainty and sensitivity study, is performed with different LBE to wall heat transfer models and different sets of the thermo physical properties of the heat exchanger material, steel and oxide layer. The investigations reveal that with best practice models the transferred power of the investigated heat exchanger design can range from 26 MW to 31 MW, with a target value of 27.5 MW. For the parametric study the thermal conductivity range of the oxide has the biggest impact on the results while for the uncertainty analysis the heat transfer model and the thermal conductivity of the oxide layer are of importance.Copyright

Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

The Karlsruhe Institute of Technology (KIT) is participating on (Code Applications and Maintenance Program) CAMP of the US Nuclear Regulatory Commission (NRC) to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test) BFBT and plant data recorded during a turbine trip event (TUSA) occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power) with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.