Risk-informed nuclear power plant LLOCA (large break loss of coolant accident) analysis method

Abstract

The invention discloses a risk-informed nuclear power plant LBLOCA (large break loss-of-coolant accident) analysis method. The method mainly comprises the following steps: 1) selecting an initial event as a nuclear power plant LLOCA; 2) according to a risk system evaluation method, establishing an event tree under the initial event, and indentifying all possible response sequences of a safety system for mitigation measures of a nuclear power plant after occurrence of the LLOCA; 3) for the event tree analysis results, combining with fault tree analysis and taking various failure data into overall consideration, quantizing occurrence possibility of all event sequences; 4) calculating peak cladding temperature corresponding to each event sequence; and 5) estimating peak cladding temperature margin of the LLOCA. By introducing a probability risk evaluation technique to a conventional deterministic analysis method, a purpose of taking overall consideration of cognitive uncertainty and accidental uncertainty of the nuclear power plant is achieved, and analysis result is closer to actual situations of the nuclear power plant.

Description

technical field [0001] The application relates to the technical field of safety analysis of nuclear power plants, in particular to a risk-guided analysis method for large breach loss of water accidents in nuclear power plants. Background technique [0002] The large break loss-of-water accident (LB LOCA) of a pressurized water reactor nuclear power plant refers to a reactor coolant loss accident caused by a large rupture in the main pipe of the reactor coolant system. The purpose of the analysis of the large breach loss of fluid accident is to verify the capacity (capacity) of the safety injection system, containment sprinkler system and auxiliary emergency water supply system (referred to as special safety facilities) of the nuclear power plant, that is, to verify the In the case of such an extreme accident, whether the reactor coolant system and special safety facilities can guarantee the integrity of the core fuel elements under various harsh environmental conditions. ...

AI Technical Summary

Problems solved by technology

[0013] To sum up, in the existing LOCA analysis methods, whether it is the traditional analysis method or the "best estimate + uncertainty analysis" method, only the cognitive uncertainty is considered, including the uncertainty of the calculation model and the Due to the uncertainty of the state parameters of nuclear power plants, conservative system assumptions are still adopted for the system, including single faults and non-safety-level systems, etc., so that the accident only develops according to a relatively conservative accident sequence, without considering the accident of the system Uncertainty, that is, the random uncertainty caused by nuclear power plant system, component failure, human error, etc. after the accident

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