Full spectrum LOCA evaluation model and analysis methodology

摘要

This invention relates to a computational system and method for performing a safety analysis of a postulated Loss of Coolant Accident in a nuclear reactor for a full spectrum of break sizes including various small, intermediate and large breaks. Further, modeling and analyzing the postulated small break, intermediate break and large break LOCAs are performed with a single computer code and a single input model properly validated against relevant experimental data. Input and physical model uncertainties are combined following a random sampling process, e.g., a direct Monte Carlo approach (ASTRUM-FS) and advanced statistical procedures are utilized to show compliance with Nuclear Regulatory Commission 10 CFR 50.46 criteria.

主权项

1. A computational system for performing a safety analysis of a postulated Loss of Coolant Accident in a nuclear reactor for a full spectrum of break sizes including large break, intermediate break and small break, comprising:
a single input model which includes noding for the full spectrum of break sizes of the postulated Loss of Coolant Accident, said input model includes an input deck which identifies parameters of the nuclear reactor selected from the group consisting of reactor geometry, initial conditions and boundary conditions, and parameters of the postulated Loss of Coolant Accident, said parameters comprising constant parameters and variable parameters; an uncertainty database including a random sampling of uncertainty values for specific parameters identifying phenomena expected to occur during the postulated Loss of Coolant Accident, a range of said specific parameters including the full spectrum of break sizes of the postulated Loss of Coolant Accident, such that a random sampling of uncertainty values is identified for the full spectrum of Loss of Coolant Accident break sizes including the large break, intermediate break and small break; an input processor to assign uncertainty values for the break size of the postulated Loss of Coolant Accident, to extract the uncertainty values from the uncertainty data base, to generate the input deck including the uncertainty values for the break size, and to execute a Loss of Coolant Accident simulation for the uncertainty values, said Loss of Coolant Accident simulation including a plurality of transient simulations executed in parallel on a cluster of processors, wherein the uncertainty values are ranged and combined following the random sampling and wherein a transfer function C(t) between a random set of values for uncertainty parameter X and a figure of merits according to peak clad temperature (PCT) and maximum local oxidation (MLO) is analyzed using the following formulation:{PCTMLO}i=1,…,N=C⁡(t)⁢{X1X2⋮Xh}i=1,…,N, where N represents a sample size; and a code to compute a response of the nuclear reactor to the Loss of Coolant Accident simulation for the uncertainty values.