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.