Uncertainty analyses of coupled thermal hydraulic/neutron kinetic code calculations for transients at NPPs with VVER reactors


Uncertainty analyses of coupled thermal hydraulic/neutron kinetic code calculations for transients at NPPs with VVER reactors

Kliem, S.; Langenbuch, S.; Weiß, F.-P.

The transition from the application of conservative models to the use of best-estimate models raises the question about the uncertainty of the obtained results. This question becomes especially important, if the best-estimate models should be used for safety analyses in the field of nuclear engineering. Different methodologies were developed to assess the uncertainty of the calculation results of computer simulation codes. One of them is the methodology developed by Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) which uses the statistical code package SUSA. In the past, this methodology was applied to the calculation results of the advanced thermal hydraulic system code ATHLET. In the frame of the recently finished EU FP5 funded research project VALCO, that methodology was extended and successfully applied to different coupled code systems, including the uncertainty analysis for neutronics. These code systems consist of a thermal hydraulic system code and a 3D neutron kinetic core model. Six different working groups applying different coupled code systems performed calculations. The involved system codes were ATHLET and SMABRE. They were used for the calculations together with the 3D neutron kinetic core models DYN3D, KIKO3D, BIPR8 and HEXTRAN.
Two real transients at NPPs with VVER-type reactors documented within the VALCO project were selected for analyses. One was the load drop of one of two turbines to house load level at the Loviisa-1 NPP (VVER-440), the second was a test with the switching-off of one of two main feed water pumps at the VVER-1000 Balakovo-4 NPP. Based on the relevant physical processes in both transients, lists of possible sources of uncertainties were compiled. They are specific for the two transients. Besides control parameters like control rod movement and thermal hydraulic parameters like secondary side pressure, mass flow rates, pressurizer sprayer and heater performance, different neutron kinetic parameters were included into the list of possible sources of uncertainties. These are the burn-up state of the core, the control rod efficiency for different control rod groups and the coefficients for Doppler and moderator density feed back. By use of the SUSA package, sets of input data with statistical variation of the relevant parameter values were generated for a large number of runs of the coupled codes for each transient.
Time-dependent rank correlation coefficients were calculated showing the influence of the varied parameters on the output parameter under investigation. The most interesting output parameters are the physical parameters for which experimental data are available. First of all, these are the core power, upper plenum pressure, core outlet and loop temperatures. The calculation results allowed also the determination of time-dependent tolerance intervals for given coverage and confidence. The comparison of the experimental data, the (best-estimate) reference solution and the tolerance intervals showed how the agreement between experiment and calculation could be quantified. In most of the cases the tolerance intervals include the experimental curves. A compiled list of the most important input parameters based on the rank correlation coefficients shows, which input parameters and models are responsible for the deviations. This list gives indications for further model improvements and code developments.

  • Lecture (Conference)
    OECD Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of LWRs - Third Workshop, 29.04.-01.05.2009, State College, USA

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Publ.-Id: 12719