Method for evaluating fuel cladding crevasse equivalent weight of nuclear power station

Abstract

The invention belongs to the technical field of operation maintenance of nuclear power stations and particularly relates to a method for evaluating the fuel cladding crevasse equivalent weight of a nuclear power station. The method comprises the following steps: firstly, an off-line sipping device is determined, fuel components are placed in a sipping room in the off-line sipping device, the sipping room is positioned on different temperature platforms through temperature rise of a water circuit, a gas fission product is released from a fuel cladding crevasse and enters a gas circuit, and a sodium iodide spectrometer performs continuous monitoring; secondly, test conditions are set; thirdly, the types of media are determined; and fourthly, the crevasse equivalent weight range of failed fuels is given according to first platform temperature, second platform temperature, the temperature rise rate and the types of the released media in combination between the correlation chart of the failed fuel crevasse equivalent weight and the Xe-133 balance time. A Fuqing nuclear power No.1 unit and a Fuqing nuclear power No.2 unit are subjected to fuel cladding failures in the first cycle, and byapplication of the method, the damaged fuel components are searched out, and the crevasse equivalent weights of damaged fuels are given out.

Description

technical field [0001] The invention belongs to the technical field of operation and maintenance of nuclear power plants, and in particular relates to an evaluation method for the crack equivalent of nuclear power plant fuel cladding. Background technique [0002] It is found from the public information that the M310 pressurized water reactor nuclear power plant using the AFA-3G fuel assembly frequently suffers from fuel cladding damage. At present, there is no clear method for judging the breach equivalent through the off-line sipping test. Therefore, it is urgent to summarize The recent off-line sipping test data of M310 pressurized water reactor, combined with the simplified interpretation guide (hereinafter referred to as S.I.G.) compiled by the French Atomic Energy Commission Cadarache Center (CEACadarache), developed and designed a nuclear power plant fuel cladding breaker. mouth equivalent evaluation method. Contents of the invention [0003] The technical problem ...

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Problems solved by technology

[0002] It is found from the public information that the M310 pressurized water reactor nuclear power plant using the AFA-3G fuel assembly frequently suffers from fuel cladding damage. At present, there is no clear method for judging the breach equivalent through the off-line sipping test. Therefore, it is urgent to summarize The recent off-line sipping test data of M310 pressurized water reactor, combined with the simplified interpretation guide (hereinafter referred to as S.I.G.) compiled by the French Atomic Energy Commission Cadarache Center (CEACadarache), developed and designed a nuclear power plant fuel cladding breaker. mouth equivalent evaluation method

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