The safety assessment of reactor thermohydraulics under different accident scenarios needs accurate prediction of conjugate heat transfer in two-phase flow. Specifically, code prediction of flow regime, cladding temperature, and heat flux in prototypic (pressures, mass and heat fluxes) steady state and transient conditions is required for accurate assessment of system margin to CHF and possibility of core damage. There is a scarce of data available and applicable for validation of numerical codes in prediction of heat transfer at high pressures and temperatures. A test campaign aiming to generate the data for validation of STH codes modelling of two-phase flow has been carried out at the Royal Institute of Technology (KTH) in Stockholm. The results are relevant to LWRs (including SMRs), and cover two-phase flow steady state conditions including approaches to CHF. The campaign was performed on High Pressure Water Test (HWAT) facility. The facility can operate at prototypic conditions in terms of pressure, temperature, flow rate and heat flux. The setup consists of a thermohydraulic loop with 3.68 m long heated section and a condenser being the ultimate heat sink. The effective height of the main loop is nine meters. The heated section is a tube with 18.9 mm inner diameter, heated using direct current. The cases tested within the campaign cover two-phase flows at pressures reaching 12.3 MPa and thermal powers up to 1.62 MW/m². The paper provides a concise literature review on two-phase heat transfer at reactor prototypic conditions, describes the experimental setup, and methodology used to calibrate GOTHIC model. Model validation is carried out focusing on approaches to critical heat flux. Conclusion on code validity and outlook for further experimental work is provided.

An experimental setup has been designed and manufactured at the Royal Institute of Technology (KTH) in Stockholm to investigate the internal structure of boiling two-phase water flow under prototypical Light Water Reactor (LWR) core conditions, including those relevant to PWR, BWR and SMR designs. The setup is based on the High-pressure WAter Test (HWAT) loop, designed for 25 MPa pressure, 1 kg/s water mass flow rate and 1 MW thermal power. The facility has been updated with a new test section and advanced instrumentation systems to enable measurements in both forced convection and natural circulation, under steady-state and transient operations. This novel experimental setup allows for the first-time measurements of radial distributions of local two-phase flow parameters under high-pressure LWR core conditions. The resulting data is intended to enhance the fundamental understanding of boiling two-phase flow phenomena, contribute to the development of closure laws (including for polydispersed flow) and support the validation of computational codes (1-D and 3-D). The paper presents the loop design, the updated instrumentation with associated uncertainties, and data post-processing methods (including the derivation of dispersed phase length scales). Results from commissioning tests, such as heat balance tests and single-phase tests, are presented. Examples of high-pressure boiling two-phase flow measurements are presented and discussed. Fundamental behavior and associated key parameters, including radial distributions of void fraction, mixture velocity, interfacial length scales and polydispersed characteristics, are identified and quantified.