Scaling in nuclear reactor system thermal-hydraulics

AP1000® Large-Break LOCA BEPU analysis with TRACE code

Comparison of global sensitivity analysis methods – Application to fuel behavior modeling

https://isiarticles.com/bundles/Article/pre/pdf/27267.pdf

The importance of input interactions in the uncertainty and sensitivity analysis of nuclear fuel behavior

Inverse uncertainty quantification of TRACE physical model parameters using sparse gird stochastic collocation surrogate model

During the recent years, an increasing interest in computational reactor safety analysis is to replace the conservative evaluation model calculations by best estimate calculations supplemented by uncertainty analysis of the code results. The evaluation of the margin to acceptance criteria, for example, the maximum fuel rod clad temperature, should be based on the upper limit of the calculated uncertainty range. Uncertainty analysis is needed if useful conclusions are to be obtained from “best estimate” thermal-hydraulic code calculations, otherwise single values of unknown accuracy would be presented for comparison with regulatory acceptance limits. Methods have been developed and presented to quantify the uncertainty of computer code results. The basic techniques proposed by GRS are presented together with applications to a large break loss of coolant accident on a reference reactor as well as on an experiment simulating containment behaviour.

/pdf/grs-method-for-uncertainty-and-sensitivity-evaluation-of-1hqka5ev3s.pdf

Horst Glaeser

Papers

GRS Method for Uncertainty and Sensitivity Evaluation of Code Results and Applications