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Increased reliance on passive emergency cooling using natural circulation of gas at elevated pressure is one of the major goals for the gas-cooled fast reactor (GFR). Since GFR cores have high power density and low thermal inertia, the decay heat removal (DHR) in depressurization accidents is a key challenge. Furthermore, due to its high surface heat flux and low velocities under natural circulation in any post-LOCA scenario, three effects impair the capability of turbulent gas flow to remove heat from the GFR core, namely: (1) acceleration effect, (2) buoyancy effect and (3) property variation. This paper reviews previous work on heat transfer mechanisms and flow characteristics of the deteriorated turbulent heat transfer (DTHT) regime and decomposes governing non-dimensional groups to provide an insight for the GFR DHR system design. It is shown that by applying the developed methodology, the GFR's DHR system has a potential for operating in the DTHT regime and the gas DTHT regime is different from the liquid or super-critical fluid's DTHT regime. A description of an experimental facility designed and built to investigate the DTHT regime is provided together with the initial test results. The initial runs were performed in the forced convection regime to verify facility operation against well-established forced convection correlations. The results of the three runs at Reynolds numbers of 6700, 8000 and 12,800 show good agreement with the Gnielinski correlation for heat transfer, which is considered the best available correlation for forced convection over a wide range of Reynolds and Prandtl numbers. However, even in the forced convection regime, the effect of variation of the fluid properties was found to be significant.

A one-dimensional single-channel thermal-hydraulics model has been developed to investigate possible occurrence of density-wave instability in two U. S. Gen-IV reactors cooled by supercritical fluids, i. e., the Supercritical Water-cooled Reactor (SCWR) and Gas-cooled Fast Reactor (GFR). Water density in the SCWR core changes from 780 kg/m3 , to 90 kg/m3 , whereas the density of supercritical carbon-dioxide in the reference GFR changes from 155 kg/m3 to around 110 kg/m3 . The standard frequency domain approach with a decay ratio of induced velocity amplitude of 0.5 has been used to determine the onset of flow instability. With suitable inlet orificing, the hot channel of SCWR has been found to be stable. Sensitivity studies show that the hot channel decay ratio reaches the critical value of 0.5 when either the reactor power is raised to 118% of full power or the core flow rate is reduced to 86% of nominal flow rate. System pressure has only a moderate effect. Detailed 3-D studies, preferably with neutronic feedback, should be carried out for the SCWR design because of its sensitivity to various important parameters. The GFR reference design has been found to be very stable since the density change in the GFR core is rather small compared to that in the SCWR design.Copyright © 2004 by ASME

The “Whole Assembly Seed and Blanket” (WASB) design, which utilizes mostly thorium in the blanket, consists of 84 seed and 109 blanket assemblies which may be backfitted into existing Pressurized Water Reactors (PWRs). Since the seed assemblies produce significantly more power than the blanket assemblies, a preliminary safety analysis of this design has been performed. Three accidents/transients (Large Break Loss of Coolant Accident (LBLOCA), Complete Loss of Primary Flow (LOPF) and Loss of Off-site Power (LOSP)), have been analyzed for both the WASB design and a typical all UO2 design for a typical 4-Loop Westinghouse PWR plant. LBLOCA results show that the peak cladding temperature (PCT) for the WASB design is approximately 260 K higher than that for a typical PWR design. However, this higher PCT for the WASB design is still about 200 K lower than the present regulatory safety limit. The response of the WASB and all UO2 core for LOPF and LOSP transients are very similar, and no post-DNB type rapid cladding temperature rise was observed in either of the two calculations.Copyright © 2004 by ASME

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.Copyright © 2004 by ASME