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Nuclear reactors are very complex units in which many physical processes occur simultaneously. Efficient heat removal from the reactor core is the most important of these processes. Heat is removed from the reactor core via heat conduction, radiation, and convection. Thus, convective heat transfer and its conditions play a crucial role in the operation and safety of nuclear reactors. Convective heat transfer in nuclear reactors is a very complex process, which is dependent on many conditions and is usually described by different correlations which combine together the most important criteria numbers, such as the Nusselt, Reynolds, and Prandtl numbers. The applicability of different correlations is limited by the conditions of heat transfer in nuclear reactors. The selection of the proper correlation is very important from the reactor design accuracy and safety points of view. The objective of this novel review is to conduct a comprehensive analysis of the models and correlations which may be applied for convective heat transfer description and modeling in various types of nuclear reactors. The authors review the most important research papers related to convective heat transfer correlations which were obtained by experimental or numerical research and applied calculations and heat transfer modeling in nuclear reactors. Special focus is placed on pressurized water reactors (PWRs), boiling water reactors (BWRs), CANDU reactors, small modular reactors (SMRs), and molten salt reactors (MSRs). For each type of studied reactor, the correlations are grouped and presented in tables with their application ranges and limitations. The review results give insights into the main research directions related to convective heat transfer in nuclear reactors and set a compendium of the correlations that can be applied by engineers and scientists focused on heat transfer in nuclear reactors. Prospective research directions are also identified and suggested to address the ongoing challenges in the heat transfer modeling of present and next-generation nuclear reactors. [ABSTRACT FROM AUTHOR]

Zirconium alloy (Zircaloy-4) is widely used for fuel cladding in the Canadian Deuterium Uranium (CANDU) heavy water reactors. However, degradation due to waterside corrosion can limit the in-reactor design life of the nuclear fuel. For this reason, efforts must be made to understand the mechanisms of corrosion and to mitigate its effects. The purpose of the experimental research related to this work consists of the assessment of corrosion of the CANDU fuel cladding material. The paper presents the results obtained by in-situ monitoring of the Pressurized Heavy Water Reactor (PHWR) water chemistry effect on fuel cladding corrosion, Zircaloy-4. Optical metallographic and Scanning Electron Microscopies (SEM), as well as X-ray diffraction (XRD) analysis, were used to evaluate the corrosion behavior of the fuel cladding material, Zircaloy-4 coupons, exposed in a Primary Heat Transport System (PHTS) autoclave system. The obtained results are useful for characterizing the corrosion behavior of Zircaloy-4 coupons exposed in the autoclave system at Cernavoda NPP for long periods. [ABSTRACT FROM AUTHOR]

A competing-rates model is presented to account for operational changes in the metastable β-Zr phase of the Zr-2.5Nb alloy used to make CANDU reactor pressure tubes and is used to predict temperature gradients at the outlet rolled joints using the decomposition of the β-Zr phase as a proxy for temperature. High temperatures decompose the β phase by enhancing the formation of small particles of ω and α phases. Fast neutron flux causes the ω and α phases to shrink. This process is assumed to depend on the total volume of the particles, because they are comparable to, or smaller than, the size of the neutron displacement cascades. The barrier energy for thermal growth was determined to be 2.43 eV, when an Arrhenius A factor of 1013/s was assumed. The cross section for (ω+α)-phase shrinkage is 24.5 barns for Zr-2.5Nb irradiated in CANDU reactors. Assuming that the shrinkage is dominated by the migration of self-interstitial point defects, a defect production efficiency of 1.4% was found. [ABSTRACT FROM AUTHOR]