Coupled Computational Heat Transfer and Reactor Physics for SCWR

2013 ◽  
Author(s):  
Christopher R. Hughes ◽  
DuWayne Schubring ◽  
Kelly A. Jordan ◽  
Dominik Rätz
Keyword(s):  
Heat Transfer ◽  
Heat Generation ◽  
Supercritical Water ◽  
Thermal Model ◽  
Computational Cost ◽  
Radial Dependence ◽  
Radial Temperature ◽  
Water Reactor ◽  
Geometry Model ◽  
Supercritical Water Reactor

To fully model the physics present within the proposed supercritical water reactor (SCWR), the thermal hydraulics calculations (yielding temperatures and densities in each material as a function of space) must be coupled to the neutronic calculations (yielding reactivity and neutron flux shapes). To enable this full coupling, a 3D model of a supercritical water reactor is being implemented in the CFD software OpenFOAM with the same geometry as a 3D MCNP (neutronics) model. Coupling will performed through result exchange between the two codes — densities and temperatures from OpenFOAM to MCNP, with heat generation returned from the neutronics calculations. Use of a reduced-geometry model is advisable due to the high computational cost of each OpenFOAM/MCNP coupling iteration. In the present work, a 1.5D thermal model of a single fuel pin was coupled with a 3D MCNP model. The thermal model includes single channel analysis (cladding/coolant heat transfer) as well as heat transfer within the cladding, helium gas gap, and uranium dioxide fuel itself. These heat transfer zones provide specific data points of the radial temperature profile. Because no radial mesh is considered, full radial dependence is not possible. The model provides limited radial dependence unlike what a 1D code could provide; thus, 1.5D is used to indicate the incomplete radial dependence that is included in the model. Iteration between the two codes is performed until heat generation is converged to within 10% between successful MCNP results. A discussion of these 3D/1.5D coupled results and the path forward to fully 3D coupling is provided.

Numerical Analysis to Investigate the Effects of Thermal-Hydraulic Instabilities on Deterioration Heat Transfer and Wall Temperature in the CANDU Supercritical Water Reactor

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Goutam Dutta ◽  
Chao Zhang ◽  
Jin Jiang
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Wall Temperature ◽  
Stability Boundary ◽  
Density Wave ◽  
Operating Regime ◽  
Maximum Temperature ◽  
Water Reactor ◽  
Heating Section ◽  
Supercritical Water Reactor

The present work analyzes the thermal-hydraulic behavior of the CANDU supercritical water reactor (SCWR) using a 1-D numerical model. The possibility of a static instability, the Ledinegg excursion, is investigated, which reveals it can occur only in a hypothetical condition, far from the proposed operating regime of the CANDU SCWR. The investigation demonstrates the possibility of density wave oscillations (DWOs), a dynamic instability, in the operating regime of the CANDU SCWR and its marginal stability boundary (MSB) is obtained. The phenomenon of the deterioration in heat transfer is observed, and the related investigation shows that the strong buoyancy effect is responsible for its appearance inside the heating section of the channel of the CANDU SCWR core. The MSB is found to be inadequate in determining the safe operating zone of the reactor because the wall temperature can exceed the allowable limit from metallurgical consideration. The investigations also determine the safe as well as stable zone where the CANDU SCWR should operate in order to avoid the maximum temperature limit and DWOs.

2D Natural Convection and Radiation Heat Transfer Simulations of a PWR Fuel Assembly Within a Constant Temperature Support Structure

2006 ◽  
Author(s):  
Pablo E. Araya Go´mez ◽  
Miles Greiner
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Fuel Assembly ◽  
Heat Generation ◽  
Temperature Difference ◽  
Wall Temperature ◽  
Radiation Heat Transfer ◽  
Radiation Heat ◽  
Fuel Rods ◽  
Water Reactor

Two-dimensional simulations of steady natural convection and radiation heat transfer for a 14×14 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 temperature was determined for a range of assembly heat generation rates and uniform basket wall temperatures, with both helium and nitrogen backfill gases. The results are compared with those from earlier simulations of a 7×7 boiling water reactor (BWR). Natural convection/radiation simulations exhibited measurably lower cladding temperatures only when nitrogen is the backfill gas and the wall temperature is below 100°C. The reduction in temperature is larger for the PWR assembly than it was for the BWR. For nitrogen backfill, a ten percent increase in the cladding emissivity (whose value is not well characterized) causes a 4.7% reduction in the maximum cladding to wall temperature difference in the PWR, compared to 4.3% in the BWR at a basket wall temperature of 400°C. Helium backfill exhibits reductions of 2.8% and 3.1% for PWR and BWR respectively. Simulations were performed in which each guide tube was replaced with four heat generating fuel rods, to give a homogeneous array. They show that the maximum cladding to wall temperature difference versus total heat generation within the assembly is not sensitive to this geometric variation.

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

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