Strategies for Developing Subchannel Capability in an Advanced System Thermalhydraulic Code: A Literature Review

Abstract

In the framework of developing next generation safety analysis tools, Canadian Nuclear Laboratories (CNL) has planned to incorporate subchannel analysis capability into its advanced system thermalhydraulic code CATHENA 4. This paper provides a literature review and an assessment of current subchannel codes. It also evaluates three code-development methods: (i) static coupling of CATHENA 4 with the subchannel code ASSERT-PV, (ii) dynamic coupling of the two codes, and (iii) fully implicit implementation for a new, standalone CATHENA 4 version with subchannel capability. Results of the review and assessment suggest that the current ASSERT-PV modules can be used as the base for the fully implicit implementation of subchannel capability in CATHENA 4, and that this option may be the most cost-effective in the long run, resulting in savings in user application and maintenance costs. In addition, improved versatility of the tool could be accomplished by the addition of new features that could be added as part of its development. The new features would improve the capabilities of the existing subchannel code in handling low, reverse, and stagnant flows often encountered in system thermalhydraulic analysis. Therefore, the method of fully implicit implementation is preliminarily recommended for further exploration. A feasibility study will be performed in an attempt to extend the present work into a preliminary development plan.

Recent Advances in Computational Convective Heat Transfer Study in a Sub-Channel for Nuclear Power Reactor and Future Directions

“…The past research also provides perspectives for future directions. In the early phases of sub-channel thermal hydraulic analysis, the focus of research articles was mostly on analytical and experimental work done for the creation of sub-channel analysis codes, [23], [24], [25], [26], [27], [28], [29], [30], [31], [32], [33], [34], [35], [36], [37], [38], [39], [40], [41], [42]. These attempts were made to increase the accuracy of CFD predictions and to derive models from CFD for novel fuel geometries, decreasing the time needed for fuel assembly design and optimization, improving safety, and running the fewest number of tests possible to conserve resources and time.…”

The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

“…They were developed for an improved description of the core thermal‐hydraulics using quasi‐3D models at component scale (decimeter to centimeter scale) for the prediction of local safety parameters such as critical heat flux (CHF) ratio, critical power ratio (CPR), fuel centerline temperature, and fuel surface temperature. A very detailed review of subchannel analysis methods and codes can be found in Moorthi et al and Cheng and Rao . In order to simulate the core thermal‐hydraulics of full cores using subchannel codes and to take profit from the increasing computer power, some subchannel codes were parallelized, eg, SCF and CORBA‐TF .…”

“…A very detailed review of subchannel analysis methods and codes can be found in Moorthi et al and Cheng and Rao. 23,24 In order to simulate the core thermal-hydraulics of full cores using subchannel codes and to take profit from the increasing computer power, some subchannel codes were parallelized, eg, SCF and CORBA-TF. 25 In addition, the codes using porous models to simulate the dedicated 3D porous media two-phase flow with a Cartesian or unstructured grid are being developed to overcome the usually 1D limitations of the system thermal-hydraulic codes.…”