A Probabilistic Safety. China Experimental Fast Reactor Design Phase Of The Internal Event Evaluation

Fast Reactor is still considered as the realistic approach to breed nuclear fuel and transmute nuclear wastes for settling the long-term energy problem. The safety of the fast reactor is always concerned by the public in all the world. The probabilistic approach, referred to more commonly as PSAs, is a reasonable supplement to the deterministic approach to the design of nuclear reactors. By the completely risk assessment of the nuclear reactor, PSA can provide a information storeroom for analysis of the specific or the universal problem. At the same time, quantitative analysis from the potential accident could assess the risk to the public and give a rather complete analysis to the safety properties of the reactor design and operation.Level I PSA is used during both the design and the operating stages of a nuclear plant to identify and analysis every possible situation and sequence of events that might result in severe core damage. Its results can therefore identify not only the weaknesses but also the strengths with regard to the reactor's safety.Based on the study of the nuclear reactor PSA and especially fast reactor PSA both in the national and international, the methodology of the Level I PSA for fast reactor is investigated and constituted. These methodology and technical include the identification and grouping of the initial events, definition of the accident sequence, modeling of the safety system, carrying through the quantitative analysis, uncertainty analysis, important and the sensitive analysis.Based on the safety design and the deterministic analysis for China Experimental Fast Reactor(CEFR), the internal initial events are established and grouped completely. Then the whole accident sequence model are defined by aid of the event tree method. The safety systems are modeled by the Failure Mode and Effective Analysis(FMEA) and the fault tree method. Then the failure data is collected from the design or the general database. Finally, the quantitative analysis including uncertainty and important/sensitive calculation is completed by the means of "Small Event Tree and Large Fault Tree" using the famous Risk Spectrum software.The result shows that the unavailability of safety shutdown system is 7.6E-10/demand, the unavailability of decay heat removal system(DHRS) is 6E-7/demand, and the total core damage frequency(CDF) is 4E-7/reactor year. The domainal initial events for the CDF is the loss of the offsite power and the domainal accident sequence for the CDF is loss of the offsite power combined with the DHRS failed.The important and the sensitive analysis shows that reducing the mission time of the DHRS, improving the reliability of the DHRS and decreasing the unavailability of the main heat transfer system(MHC) by optimizing operation rules and perfect safety management are the main aspects to improve the reliability of the CEFR.This study is the first probabilistic assessment of the fast reactor in China. A set ofapplicable PSA methodology for fast reactor is investigated and practiced. It will be used in the deeply detail analysis in future.It is the first study to actualize the PSA for CEFR and get the quantitative reliability parameters. These results shows according with the design targets for fast reactor advised by IAEA and nearly to that of the other fast reactor PSAs. Also it is new confirmation to that the CDF of the fast reactor is normally smaller about 10 or 100 times compared with PWR. And it proved the safety of the fast reactor as the import part of the China nuclear fuel cycle in future.GO method, a special method for the analysis of system, is innovatively used to the modeling of the power system of CEFR in the study. The theory model of the CEFR power system based on the GO principle is founded and a computer code is developed to finish the calculation. The results from this analysis is used in the safety system quantitative analysis.