Nuclear data sensitivity for reactor physics parameters in a lead-cooled reactor

The purpose of this research is to study and investigate the evaporation lifetime and boiling curve on hemispherical heated surface. The selected material was stainless steel (304). A nearly perfect and smooth hemispherical surface was developed by using EDM die sinker. For the test liquid, distilled water was used during the experimental work. The average droplet temperature was 31.36 °C corresponding to liquid subcooling ΔT sub = 68.64K. Based on the theoretical calculation, the diameter of the water droplet was approximately 5.00 mm. Meanwhile, the impact height was approximately 65.0 mm corresponding to the theoretical impact velocity of 1.129 m/s. The material was heated using a digital hot plate which was able to give an accurate reading and stable temperature fluctuation during the heating process. The temperature ranged from a low temperature of T w = 100 °C to a high temperature of T w = 300 °C. As a result, the boiling curve showed a similar pattern of other experimental work that consists of two (2) important points which are Critical Heat Flux(CHF) and Leidenfrost temperature.

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Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison

Chereshkov,

Ternovykh,

Tikhomirov

et al. 2023

NUCET

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The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.

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Nuclear data sensitivity for reactor physics parameters in a lead-cooled reactor

Cited by 3 publications

References 38 publications

Negative temperature coefficients of reactivity for metallic fuel Dual Fluid Reactor

Negative temperature coefficients of reactivity for metallic fuel Dual Fluid Reactor

Evaporation lifetime and boiling curve on hemispherical stainless steel (304) surface

Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison

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Nuclear data sensitivity for reactor physics parameters in a lead-cooled reactor

Cited by 3 publications

References 38 publications

Negative temperature coefficients of reactivity for metallic fuel Dual Fluid Reactor

Negative temperature coefficients of reactivity for metallic fuel Dual Fluid Reactor

Evaporation lifetime and boiling curve on hemispherical stainless steel (304) surface

﻿Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison

Contact Info

Product

Resources

About

Nuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison