He jishu (Oct 2023)
Systematic measurement method for primary coolant flow rate in advanced passive nuclear power units
Abstract
Background The primary coolant flow rate is essential in preventing departure from nucleate boiling. The implementation of a low-leakage core loading pattern in advanced passive (AP) technology-based nuclear power units has increased the temperature difference gradient at the core outlet, resulting in elevated uncertainty in the flow rate calculations when using the heat balance method. Purpose This study aims to validate a measurement and calculation method based on the Bernoulli equation model for accurately determining the primary coolant flow rate in AP nuclear power units, hence meeting the design and regulatory requirements. Methods First of all, measurements were conducted for the primary loop main equipment and bend pipe flowmeter pressure differentials during the commissioning phases. Calorimertic balance tests were performed at power levels of 50%, 75%, 90%, and 100%. Then, the bend pipe flowmeter coefficients were calibrated using the flow rate values obtained from the hot function test and 100% rated thermal power (RTP). Finally, based on weighted factors, the total flow rate values for the reactor coolant system (RCS) were calculated with emphasis on the minimization of uncertainties. Results The proposed measurement and calculation method yields primary coolant flow rate values with a relative error of less than 4%. The total flow rate after loading is within the range of 95.8% to 104% of the expected optimum flow rate. The uncertainty of the volumetric flow rate calculated from NAPs is lower than 1.9%, demonstrating a novel approach for precise measurements in other units. Conclusions The method of this study offers an advanced perspective for reactor coolant precise measurements in other units, with primary coolant flow rate values exhibiting minimal relative error and volumetric flow rate values from NAPs demonstrating low uncertainty.
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