Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

Gerhard Strydom

doi:10.1155/2013/426356

Abstract

The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transient PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.

Highlights

E time behavior of the maximum fuel temperature during the depressurized loss of forced cooling (DLOFC) transient is shown in Figures 4, 5, and 6 for the rst 30 cases of the 100 Simple Random Sampling (SRS) Uniform, 100 Latin Hypercube Sampling (LHS) Normal, and 200 LHS Normal datasets, respectively
(i) e PBMR core design leads to the typical High Temperature Gas Cooled Reactors (HTGRs) loss of cooling behavior, that is, a slow increase in the maximum fuel temperature over several hours, with the peak fuel temperature reached 40–60 hours into the transient
The same physical phenomena are present in all the DLOFC events, the rate of energy deposition and energy removal differ for each of these cases, according to the sampled input values

Summary

Introduction

Title 10 Part 50 (10 CFR 50.46) of the United States Code of Federal Regulations rst allowed Best Estimate calculations rather than conservative code models of safety parameters in nuclear power plants in the 1980s, stipulating, that uncertainties be identi ed and quanti ed 1]. e continued development of High Temperature Gas Cooled Reactors (HTGRs) requires validation and veri cation of HTGR design and safety models and codes, and the predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. In many complex uid system simulation codes, the combination of governing equations and subgrid correlations yields ill-posed systems of differential equations that do not converge to the analytical solution upon re nement of the mesh Another important source of uncertainty is that the input model is a simpli cation of the actual physical geometry. Several large international uncertainty quanti cation programs have been developed in recent years, of which the LWR Uncertainty Analysis in Modeling (UAM) benchmark [2] and the earlier BEMUSE [3] project are two examples Both these efforts are coordinated by the Organization for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA). A dedicated OECD/NEA code-to-code benchmark on the MHTGR-350 design was recently launched [8], and the same codes and models will be utilized in the core modeling part of the HTGR UAM study

Uncertainty Methodology

Discussion of PEBBED-THERMIX Results

SUSA Uncertainty Analysis

Findings

Conclusions