Abstract
A comprehensive uncertainty analysis in the event of a severe accident in a two-loop pressurized water reactor is conducted using an uncertainty package integrated in the ASYST code. The plant model is based on the nuclear power plant (NPP) Krško, a Westinghouse-type power plant. The station blackout scenario with a small break loss of coolant accident is analyzed, and all processes of the in-vessel phase are covered. A best estimate plus uncertainty (BEPU) methodology with probabilistic propagation of input uncertainty is used. The uncertain parameters are selected based on their impact on the safety criteria, the operation of the NPP safety systems and to describe uncertainties in the initial and boundary conditions. The number of required calculations is determined by the Wilks formula from the desired percentile and confidence level, and the values of the uncertain parameters are randomly sampled according to appropriate distribution functions. Results showing the thermal hydraulic behaviour of the primary system and the propagation of core degradation are presented for 124 successful calculations, which is the minimum number of required calculations to estimate a 95/95 tolerance limit at the 3rd order of the Wilks formula application. A statistical analysis of the dispersion of results is performed afterwards. Calculation of the influence measures shows a strong correlation between the decay heat and the representative output quantities, which are the mass of hydrogen produced during the oxidation and the height of molten material in the lower head. As the decay heat increases, an increase in the production of hydrogen and the amount of molten material is clearly observed. The correlation is weak for other input uncertain parameters representing physical phenomena, initial and boundary conditions. The influence of the order of the Wilks formula is investigated and it is found that increasing the number of calculations does not significantly change the bounding values or the distribution of results for this particular application.
Highlights
Uncertainties in assessing the nuclear power plant (NPP) behaviour according to predictions of severe accident codes affect the on-site and off-site risk assessment
Themass molten pool will todecreases the lowercontinuously head if the crust the periphery of the core that this decrease is faster when the pressure in the primary system is the higher, is surrounding the corium fails, which takes place when the stresses in crustwhich become expected because the higher the pressure, therelocation greater the forcebetween acting on the sfluid the higher than its ultimate strength
A comprehensive uncertainty analysis of a severe accident requires the selection of uncertain parameters that cover both thermohydraulic and core damage phenomena, their probability density functions and qualified best estimate plus uncertainty (BEPU) methodology
Summary
Uncertainties in assessing the nuclear power plant (NPP) behaviour according to predictions of severe accident codes affect the on-site and off-site risk assessment. In the area of severe accidents there are fewer examples of the use of uncertainty quantification, and these include estimating the influence of uncertainty of input parameters and assumptions on fission product releases [5–7] and hydrogen production [8] These analyses cover a station blackout sequence at the U.S NPP Sequoyah, Westinghouse. THA module for thermal hydraulic calculation that uses best estimate two-fluid, nonequilibrium models and correlations, the SCDAPSIM module for calculation of reactor core severe accident progression and the SCDAPSIM COUPLE module for the finite element thermal analysis of the reactor pressure vessel lower head It uses a variety of other member-developed computational packages. The uncertainty package integrated in the ASYST code is based on the BEPU methodology with probabilistic propagation of input uncertainty [32] It consists of the following steps: Development of the nuclear power plant computation model for the steady state and the transient calculation; ing steps: Energies 2022, 15, 1842. Other important output results are presented and commented on
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