VERIFICATION, VALIDATION AND UNCERTAINTY QUANTIFICATION OF A NUCLEAR THERMAL HYDRAULICS CODE

Nuclear thermal hydraulics codes are used in designing next generation systems and analyzing existing designs. Since most of the nuclear safety analyses employ nuclear thermal hydraulics codes, every step of the development in these codes are carefully verified and validated (V&V). This study shows the V&V steps of uncertainty equations implemented into the nuclear safety code of Coolant Boiling in Rod Arrays Code-Two-Fluid (COBRA-TF). COBRA-TF, designed by Pacific Northwest Laboratory, represents a two-fluid, three-field (continuous liquid, continuous vapor and entrained liquid drop) representation of two-phase flow. For heat transfer from and within the solid structures in contact with the fluid, a finite difference and semi-implicit numerical technique on an Eulerian mesh is used to solve conservation equations for each of the three fields. Even though the code is capable of predicting two-phase flow response of a system, it only predicts deterministic results without uncertainty bounds. Therefore, uncertainty equations based on Aydogan's sampling uncertainty method are implemented into COBRA-TF to obtain uncertainty bounds of code predictions. The V&V steps of US-NRC's Regulatory Guide 1.203 (Rg 1.203) are followed as a guideline after the code updates. Several code-to-data comparisons are done in the process of V&V: single phase pressure drop, two phase pressure drop, void distribution, critical power and dry-out location. Uncertainty bounds of code predictions are calculated and compared with the experimental uncertainty bounds. An experimental database which covers various two phase flow experiments, boundary conditions (mass flow rate, pressure and inlet enthalpies), 1/1 scale nuclear fuel bundles, axial and radial power distribution is selected for the purpose of this study. The uncertainty results of new uncertainty equations coded in COBRA-TF are satisfying.