6 results about "Sodium-cooled fast reactor" patented technology

The invention provides a gas concentration time sequence distribution prediction method based on multivariable data driving. Based on the theoretical relationship between the mean square displacement and the mean scattering angle cosine of neutrons in an infinite homogeneous medium, the mean square displacement is directly counted through Monte Carlo simulation, the mean scattering angle cosine is reversely deduced, and a first-order scattering matrix meeting the mean square displacement conservation is constructed accordingly. In order to be suitable for a finite geometric model, a correction factor is further introduced to correct an average scattering angle cosine obtained by a traditional method. A verification result shows that the method remarkably reduces the calculation deviation of effective proliferation factors, improves the neutron flux distribution precision, and is particularly suitable for high-precision multi-group calculation of high-anisotropy fast spectrum reactor cores such as pebble-bed high-temperature gas cooled reactors and sodium cooled fast reactors. According to the method, the time sequence probability distribution prediction of the gas concentration can be realized, the prediction uncertainty is quantified, the risk indexes such as the over-limit probability are directly output, and the scientificity and the reliability of coal mine safety early warning are improved.

In a sodium-cooled fast reactor, a breached fuel pin releases fission and activation products into the primary sodium coolant and into the cover gas. A portion of the primary sodium coolant is diverted to a distillation system where its constituents can be separated based on volatility. Condensing these constituents enables precise measurement of their concentration without delays caused by waiting for the decay of radioactive sodium background scatter, enhancing the detection and quantification of the dilute constituents. Additionally, the method continuously determines the initial concentration of constituents in the feed. Quantification is achieved using detectors, facilitating the continuous assessment of dilute constituent concentrations in the sodium coolant feed stream.

A nuclear reactor is designed and constructed to withstand seismic events. A mechanical hold down and discriminator may be provided as cooperating structure between the core receptacle and a nozzle of a core assembly. One or more receivers may be formed on the nozzle that cooperate with corresponding protrusions in a core receptacle. As the nozzle is inserted into the receptacle, the protrusion engages with the receiver to inhibit unintentional withdrawal of the nozzle from the receiver. A first pattern of receivers on a nozzle may align with a first pattern of protrusions within a receptacle to enable the nozzle to fit within a first receptacle. A second pattern of receivers formed on a second nozzle may inhibit the second nozzle from being inserted into the first receptacle and thereby provide a discrimination feature to ensure proper core assemblies are located in the correct location within a nuclear reactor core.

A sodium-air heat exchanger according to the present disclosure may comprise: a vertically erected tubular body through which air flows; a heat transfer tube bundle arranged longitudinally within the tubular body through which sodium flows; an upper duct installed at the top of the tubular body through which sodium is introduced and air is discharged; and a lower duct installed at the bottom of the tubular body through which sodium is discharged and air is introduced.

The embodiment of the invention relates to the field of nuclear reactor monitoring, in particular to a method and system for detecting leakage of a sodium-cooled fast reactor steam generator. The method for detecting leakage of the sodium-cooled fast reactor steam generator comprises the following steps: S1, collecting vibration data of a steam generator shell; s2, performing gain and filtering processing on the vibration data; s3, acquiring a power spectrum corresponding to the vibration data; and S4, determining that the steam generator leaks according to the power spectrum. Compared with the method for determining the leakage of the steam generator according to the measured hydrogen concentration, the method for detecting the leakage of the sodium-cooled fast reactor steam generator provided by the embodiment of the invention has the advantages that the vibration data of the shell of the steam generator are collected and processed to determine the leakage of the steam generator; therefore, the leakage of the steam generator can be quickly detected, so that online monitoring of the leakage of the steam generator can be realized, and the safety of the steam generator can be ensured.