The supercritical water-cooled reactor (SCWR) is essentially a pressurized water reactor operating above the thermodynamic critical point of water (T_c=647.096K, P_c=22.064 MPa). It is considered as one of the most promising Generation IV reactors because of its simplicity, high thermal efficiency, and nearly fifty years of industrial experience from thermal-power stations with a SCW cycle. Evolving from the existing designs, there are currently two types of SCWR concepts: (a) a large reactor pressure vessel containing the reactor core (fuelled) heat source, analogous to conventional PWRs and BWRs, and (b) distributed pressure tubes or channels containing fuel bundles, analogous to conventional CANDU and RBMK nuclear reactors.

The supercritical water coolant remains in single phase at all operating conditions of the SCWR. Therefore, the traditional limiting criterion, based on critical heat flux phenomenon, is not applicable. In turn, maximum cladding surface temperature (MCST) and peak fuel centerline temperature have been adopted as design criterion for the SCWR. Similar to existing subchannel-code analyses of conventional reactor fuels, the cladding temperature depends mainly on the flow distribution, accuracy of empirical equations (such as heat transfer, turbulent mixing, and hydraulic resistance). The challenge envisioned in the analysis is the availability of closure relationships for supercritical water applications.

A subchannel code (ATHAS, Advanced Thermal-Hydraulics Analysis Subchannel) is developed to meet the demand of the preliminary analyses of flow and enthalpy distributions and fuel cladding temperature at supercritical water conditions and of the optimum design of fuel bundle. The code is based on the basic transient mass, momentum and energy equations. Due to the lack of relevant experimental data for subchannel parameters, a literature survey of heat transfer, hydraulic resistance, and turbulent mixing at supercritical pressures has been performed to compile applicable correlations for implementation into the code. A total of 13 heat transfer correlations, 6 frictional resistance correlations, and 13 mixing models have been collected and implemented into the code as options for sensitivity analyses. A 3-dimensional fuel conduction model has been implemented into ATHAS to consider the highly heterogeneity of the SCWR bundle. The calculation of supercritical water properties is based on a combined method of look up table and correlations, improving the code performance and accurate. The code is applicable to different bundles types, such as the hexagonal arrangement and square arrangement with moderator water rod currently used in the pressure-vessel reactors, and circular-ring arrangement currently adopted in the pressure-tube reactors. A cross-code comparison has been preformed whereas the predictions of ATHAS are compared against those from (1) the STAFAS code for the HPLWR fuel design; (2) the VIPRE code for the fuel design with square arrangement of fuel rods and moderator water rod; and (3) the Japanese subchannel code with supercritical fast reactor code. ATHAS has been shown applicable to analyses of these fuel designs with similar results to those predicted using other codes. A sensitivity analysis has been performed to assess various thermalhydraulics correlations and the 3-D fuel model on cladding temperature predictions for the 43-rod CANDU type and RBMK type bundles.

The work was funded by AECL (Canada), and XJTU, CGNPC and NPIC (China).