The experimental investigations were carried out of a pressurized water nuclear reactor (PWR) with enhanced surface using different concentration (0.5 and 2.0 vol%) of ZnO/DI-water based nanofluids as a coolant. The experimental setup consisted of a flow loop with a nuclear fuel rod section that was heated by electrical current. The fuel rod surfaces were termed as two-dimensional surface roughness (square transverse ribbed surface) and three-dimensional surface roughness (diamond shaped blocks). The variation in temperature of nuclear fuel rod was measured along the length of a specified section. Heat transfer coefficient was calculated by measuring heat flux and temperature differences between surface and bulk fluid. The experimental results of nanofluids were compared with the coolant as a DI-water data. The maximum heat transfer coefficient enhancement was achieved 33% at Re = 1.15 × 105 for fuel rod with three-dimensional surface roughness using 2.0 vol% nanofluids compared to DI-water.
The applicability of artificial roughness in light-water reactors is investigated for the purpose of heat transfer improvement in fuel rod bundles. Since the roughening technique has a significant impact on friction losses, the investigation is divided in two distinct steps: flow resistance and convective heat transfer. The present paper deals with roughness effects on flow resistance. The technique consists of a multiplicity of small elements distributed on the surface of the simulated fuel rod. A parallel rib-type roughness is selected for this work for simplicity and since it has been extensively investigated in the past. Locally flow resistance is simulated using Computational Fluid Dynamics, CFD, in smooth and in rough rod bundles downstream of support grids with and without flow-enhancing features (vanes). This investigation is performed with basis on experimental testing. With model parameters established, various candidate roughness designs can evaluated for minimum flow resistance.
The effects of artificial roughness for the purpose of thermal performance improvement in pressurized water nuclear reactors are investigated. The artificial roughness consists of two-dimensional ribs parallel to the turbulent flow. The fuel rod bundle subchannel is preliminarily modeled as an annulus using the finite element method in ANSYS/FLOTRAN. The Navier-Stokes equations are solved from the SST (Shear Stress Transport) turbulence model for the simulated annulus thermal-flow. The analyses are performed for ribs dimensions and pitch provided by published previous work. It is found that, heat transfer and differential pressure have similar behavior with highest heat transfer occurring at the reattachment point. The finite element model describes well the characteristics of turbulent flow in smooth and rough rod when compared to previous semi-empirical models. Next paper extends the analysis by comparing numerical results with experimental test data and sensitivity analyses for different roughness configurations.
This paper presents convective heat transfer results for enhanced surface rod bundles for a pressurized water nuclear reactor (PWR). The experiments, to determine the single-phase heat transfer characteristics, were performed in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for PWR. In this experimental investigation, the effects of enhancing the outer surface roughness of a simulated nuclear fuel rod bundle were studied. The outer surface of the simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The simulated rod was made of Inconel 615, and was supported by Nickel 200 rods. Joule heating was supplied to the rod by passing electric current through the rod. The rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm, placed in a test-section made of 38.1 mm inner diameter aluminum pipe with a wall thickness of 6.35 mm. Tests were conducted for Reynolds numbers ranging from 105 to 106. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer coefficient enhancement recorded was 86% at Re = 4.18×105.
This work aims to develop a finite element model of a PWR control rod at operating conditions for stress analysis of the rod cladding. The finite element model simulates a control rod exposed to high operating temperatures and pressure while portions of the rod are irradiated, resulting in accumulated fluence of neutrons by the rod materials. These high temperature and accumulated fluence induce thermal expansion and swelling of the rod materials, especially of the absorber, which may eventually interact with the rod cladding, generating stresses and strains in the wall of the cladding tube. Moreover, if the maximum stress or strain in the tube wall exceeds the design allowable limit, the absorber rod is considered failed. The author creates the control rod finite element model and apply the operating loads on two-dimensional axisymmetric elements to obtain displacements, temperatures, stresses, and strains. The model also includes contact surface elements to evaluate eventual mechanical interactions between absorber and cladding due to thermal expansion and swelling effects. This is a coupled nonlinear static analysis solution that includes thermal expansion effects to calculate temperature distribution and subsequent thermal strains in the absorber rod due to the heat generation rates and coolant temperature; swelling analysis to calculate absorber growth induced by irradiation; and creep analysis to calculate absorber stress relaxation under coolant pressure and temperature. The finite element model is capable of determining whether or not absorber-to-cladding gap closure will occur and if so, calculate maximum stress and strain in the rod cladding associated with mechanical interaction between the two components induced by the operating temperature and thermal fluence loads.
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