Following the severe accidents at the Japanese Fukushima Daiichi Nuclear Power Station in 2011, the US Department of Energy initiated research and development on the enhancement of the accident tolerance of light water reactors by the development of fuels/cladding that, in comparison with the standard UO2/ Zircaloy (Zr) system, can tolerate loss of active cooling in the core for a considerably ...

Abstract—TRACE is developed by U.S. NRC for the nuclear power plants (NPPs) safety analysis. We focus on the establishment and application of TRACE/FRAPTRAN/SNAP models for Chinshan NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17 m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are three TRACE/SNAP models: one-channel, two-channel, and multi-channel TRACE/S...

The thermophysical performance and solid mechanics behavior of UO2-36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition, the effect of coating thickness (0.5, 1 and 1.5 mm) on fuel performance and mechanical interaction is discussed. The...

Link for citation: Rofida Hamad Khalifa Hamad, Nikitenkov N.N., Kudiyarov V.N. Calculation of nuclear fuel cladding internal coating thickness emergency resistant element. Bulletin the Tomsk Polytechnic University. Geo Аssets Engineering, 2023, vol. 334, no. 6, рр.27-34. In Rus. The relevance study lies in need to create an reactors, based on consequences accidents Chernobyl and Fukushima. spec...

Abstract—Analysis of the uncertainty quantification related to nuclear safety margins applied to the nuclear reactor is an important concept to prevent future radioactive accidents. The nuclear fuel performance code may involve the tolerance level determined by traditional deterministic models producing acceptable results at burn cycles under 62 GWd/MTU. The behavior of nuclear fuel can simulat...

Fuel-Cladding Chemical Interaction (FCCI) in metallic nuclear fuels is a key phenomenon that limits the maximum burn-up and operating power as it can cause premature failure of fuel cladding. Therefore, being able to accurately predict interactions fuel-cladding system by means thermodynamic calculations contributes our understanding how mitigate this mechanism. In order test capabilities Therm...

A two-dimensional computational model of a spent 7x7 Boiling Water Reactor assembly in a horizontal support basket was developed using the Fluent computational fluid dynamics package. Heat transfer simulations were performed to predict the maximum cladding temperature for assembly heat generation rates between 100 and 600W, uniform basket wall temperatures of 25 and 4000C, and with helium and n...

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14x14 pressurized water reactor (PWR) spent nuclear fuel assembly within a square basket tube of a typical transport package were conducted using a commercial computational fluid dynamics package. The assembly is composed of 176 heat generating fuel rods and 5 larger guide tubes. The maximum cladding temp...