Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

Abstract

In this study, uncertainty and sensitivity analyses were performed with MELCOR 2.2.18 to study the hydrogen generation (figure-of-merit (FoM)) during the in-vessel phase of a severe accident in a light water reactor. The focus of this work was laid on a large generation-III pressurized water reactor (PWR) and a double-ended hot leg (HL) large break loss of coolant accident (LB-LOCA) without a safety injection (SI). The FPT-1 Phebus integral experiment emulating LOCA was studied, where the experiment outcomes were applied for the plant scale modelling. The best estimate calculations were supplemented with an uncertainty analysis (UA) based on 400 input-decks and Latin hypercube sampling (LHS). Additionally, the sensitivity analysis (SA) utilizing the linear regression and linear and rank correlation coefficients was performed. The study was prepared with a new open-source MELCOR sensitivity and uncertainty tool (MelSUA), which was supplemented with this work. The FPT-1 best-estimate model results were within the 10% experimental uncertainty band for the final FoM. It was shown that the hydrogen generation uncertainties in PWR were similar to the FPT-1, with the 95% percentile being covered inside a ~50% band and the 50% percentile inside a ~25% band around the FoM median. Two different power profiles for PWR were compared, indicating its impact on the uncertainty but also on the sensitivity results. Despite a similar setup, different uncertainty parameters impacted FoM, showing the difference between scales but also a significant impact of boundary conditions on the sensitivity analysis.

Uncertainty Study of the In-Vessel Phase of a Severe Accident in a Pressurized Water Reactor

“…A significant experience exists in the application of best estimate thermal hydraulic system codes for quantification of uncertainty during different types of calculations: the analyses of a large break loss of coolant accident (LOCA) in experimental facilities [1] and NPPs [2,3], assessment of thermal hydraulic phenomena in boiling water reactors with a focus on plant systems and fuel performance, two-phase flow and sub channel analysis [4]. 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][6][7] and hydrogen production [8]. These analyses cover a station blackout sequence at the U.S. NPP Sequoyah, Westinghouse 4-loop pressurized water reactor (PWR) [5], 20 accident sequences covering a wide range of conditions in eight reactor designs that are operated in the European Union [6] and the analyses of experiment PHEBUS FPT-1 conducted in the PHEBUS facility used to investigate reactor accident scenarios in PWR reactors [7,8].…”

“…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][6][7] and hydrogen production [8]. These analyses cover a station blackout sequence at the U.S. NPP Sequoyah, Westinghouse 4-loop pressurized water reactor (PWR) [5], 20 accident sequences covering a wide range of conditions in eight reactor designs that are operated in the European Union [6] and the analyses of experiment PHEBUS FPT-1 conducted in the PHEBUS facility used to investigate reactor accident scenarios in PWR reactors [7,8]. Severe accident codes have reached a high level of development and are used, among other applications, to support severe accident management programs [9,10] and to assess modifications in nuclear power plants [11].…”