Doctoral Dissertations

Tanju Sofu

Date of Award

12-1992

Degree Type

Dissertation

Degree Name

Doctor of Philosophy

Major

Nuclear Engineering

Major Professor

H. L. Dodds

Committee Members

T. W. Kerlin, B. R. Upadhyaya, A. E. Ruggles, E. L. Wachspress, J. Smith, D. H. Cook

Abstract

A coupled neutronics and core thermal-hydraulic performance model has been developed for the analysis of high flux research reactor reactivity insertion accidents. The model is developed specifically for the ORNL's High Flux Isotope Reactor (HFIR) but can be adapted to other aluminum clad, plate-type core geometries. The model includes point reactor kinetics for neutronics, a one-dimensional, non-homogeneous, equilibrium two-phase flow and heat transfer model with provision for subcooled boiling in the coolant channels, and a spatially averaged, one-dimensional heat conduction model for the fuel plates. In addition, the feedback from core regions other than the fuel elements is included by employing a lumped parameter approach with a one node representation of the metal and a two-node representation of the coolant in these non-fueled regions. The safety system response is incorporated in the model by approximating the movement of the safety quadrants with a second order differential equation. This basic model is used to describe the transient behavior for both nominal and hot-spot conditions of the inner and outer fuel elements, both at the beginning and at the end of the fuel cycle.

A direct numerical solution of the multiple point difference approximation to the equations representing the core dynamics model have been used in an implicit fashion. In this scheme, the partial differential equations of the model are discretized in space and then combined with the rest of the dynamic equation set. This leads to a large initial value problem of ordinary differential equations. A variable-order, variable-time-step time- advancement-scheme based on the forward and backward Euler method is used to solve this initial value problem.

The model is used to analyze a series of HFIR reactivity transients associated with malfunction of the reactor control system, formation of voids in the regions with positive reactivity coefficients, and pump start in a cold loop. Each transient is studied for four operating states of the reactor which are characterized by distinct initial conditions and safety system settings. The results of this analysis indicate that, in addition to the unprotected sequence for uncontrolled shim withdrawals, the worst credible reactivity accident for the HFIR is due to optimum amount of voids in the target region. For this case, a partial hot- plate melting is predicted in Mode 1, at full power, and at the end of cycle conditions.

The analysis is supplemented with a study for the consequences of local hot-plate melting by examining the initial disruption modes of the core at the onset of melting. Scoping calculations are performed to assess the potential for molten fuel entrainment at the expected hot-channel conditions. The potential consequences of partial flow blockages are also addressed by examining the dynamics of molten fuel accumulation as it moves into the subcooled water at the channel outlet and then cools. It is concluded that the fuel melting will be limited to the hot-plate and can be tolerated without challenging the overall mechanical integrity of the rest of the core. In summary, the HFIR can withstand the maximum credible reactivity accident with only a small amount of fuel melting.

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