Recent challenges in subchannel thermal-hydraulics-CFD modeling, subchannel analysis, CHF experiments, and CHF prediction

• Published in: Nuclear engineering and design Vol. 354; p. 110236
• Main Authors: Yang, Bao-Wen, Han, Bin, Liu, Aiguo, Wang, Sipeng
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 01.12.2019; Elsevier BV
• Subjects: Assemblies; Boiling; Bundling; Computational fluid dynamics; Computer applications; Cross flow; Experiments; Fluid dynamics; Fluid flow; Heat flux; Hydraulics; Hydrodynamics; Light water reactors; Mathematical models; Nuclear accidents & safety; Nuclear energy; Nuclear engineering; Nuclear fuels; Nuclear power plants; Nuclear reactors; Nuclear safety; Nucleation; Open channels; Pressurized water reactors; Reactors; Safety
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2019.110236

•For the first time, the hidden non-conservatisms due to non-typical CHF physical phenomena and non-representative exit quenching effects for both uniformly heated and potentially exit peaking power profile rod bundle CHF tests were identified. These effects not only observed in the typical (straight) rod bundle but also in the bowed rod bundle CHF tests using either axial uniform heater rods or exit peaking (or top skewed) power profile.•The issue of non-representation in using Freon as a surrogate for water in rod bundle CHF tests due to non-compatible scaling was first explored.•A new CHF mechanistic model, the Non-uniform heater Homogeneous Nucleation Model (NHNM), was introduced along with a proposal for new or modified boiling curve due to a jump from a point between ONB and OSV to transition.•A new application with Distributed Resistance Model (DRM) was also first proposed for rod bundle fuel assemblies with mixing vane grids to account for the contribution of cross flow mixing in spacer grids using subchannel analysis codes. Reactor core thermal hydraulics is, or has always been, one of the key components for the safety of nuclear reactor. Continuous efforts have been devoted to the investigation and understanding of its basic physical phenomena and safety analysis. Due to its complex geometry, open channel interactions, non-uniform axial/radial heating, periodic mixing vane spacer grid (MVG) mixing (Yang et al., 2014a), and broad range of parameters, the rod bundle subchannel thermal-hydraulics in light water reactors (LWR) is, in particular, challenging for both modeling and experimental investigations. Recently, the subjects of reactor core thermal-hydraulics have gained more attention in the international community through three international seminars (ISACC 2013 Xian China, ISACC 2015 Shenzhen China, and IS-ReCTHA 2018 Lake Lecco, Italy) and several journal issues (STNI 2014 [Yang et al., 2014b], Kerntechnik 2016 [Yang et al., 2016], and NED 2019 [Ninokata et al., 2018]) on reactor core thermal-hydraulics with special focuses on Computational Fluid Dynamics (CFD) modeling, subchannel analysis, and rod bundleCritical Heat Flux (CHF) experiments. As a part of a series of reviews, this paper presents a brief summary of some ongoing key issues concerning various aspects of CFD modeling, subchannel analysis, rod bundle experiments, and rod bundle CHF modeling and prediction that are critical for understanding the underlying fundamental physical phenomena, the advancement in reactor core thermal-hydraulics, and its safety applications for commercial reactors and nuclear power plants. Not only were various challenges in rod bundle CHF and CFD modeling presented in this paper, but for the first time after over 30 years of practices, the hidden non-conservatisms due to non-typical CHF physical phenomena and non-representative exit quenching effects for both uniformly heated and potentially exit peaking power profile rod bundle CHF tests were also identified (Yang et al., 2014a). Another topic recently brought to light is the issue of non-representation in using Freon as a surrogate for water in rod bundle CHF tests due to non-compatible scaling (Yang et al., 2014a). In particular consideration for the non-uniform high axial power peaking characteristics associated with PWR fuel, especially in short bundle fuel assemblies, a new CHF mechanistic model, the Non-uniform heater Homogeneous Nucleation Model (NHNM), was introduced along with a proposal for new or modified boiling curve due to a jump from a point between ONB and OSV to transition boiling or boiling crisis. A new application with Distributed Resistance Model (DRM) was also first proposed (Mao and Yang et al., 2017) for rod bundle fuel assemblies with mixing vane grids to account for the contribution of cross flow mixing in spacer grids using subchannel analysis codes.