Honors College Thesis

 

An Examination of Critical Heat Flux in a Nuclear Fuel Rod Using an Explicit Finite Difference Scheme Public Deposited

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https://ir.library.oregonstate.edu/concern/honors_college_theses/pk02cg51d

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  • One of the most important components in the safety analysis of a nuclear reactor is its critical heat flux (CHF), as it can compromise the structural integrity of the clad and lead to the release of fission products into the primary coolant. Groeneveld et. al. has published a series of look-up tables used in the prediction of the onset of CHF. A forward-element finite difference method was written to model the temperature distribution in a fuel rod and, using a series of look-up tables published by Groeneveld et. al, the CHF and critical heat flux ratio determined and compared against that previously published for a Westinghouse PWR. Two test cases were run using the finite difference method, one with a uniform volumetric heat generation distribution, the other a sinusoidal distribution. Included in the analysis of the sinusoidal distribution are hot spot factors associated with typical LWRs. While the preliminary steady state test results appear accurate, future tests need to be performed using transient power distributions. Additionally, the assumption of no axial heat transfer significantly affects the critical heat flux ratio (CHFR) at nodes near the inlet and outlet of the core when using a sinusoidal heat generation profile. This assumption should be removed prior to the conduction of transient tests.
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