In Japan Atomic Energy Agency, an in-house subchannel analysis code called ASFRE have been developed to evaluate fuel assembly (FA) thermal-hydraulics in sodium-cooled fast reactors (SFRs). In the FA, the hexagonal wrapper tube is equipped with fuel rods which is spirally wound by wire-spacer to maintain the sodium coolant flow area between the fuel rods and to promote the mixing of coolant and enhance the heat transfer. For the safety design of FA, use of high-reliability numerical analysis codes is required to evaluate the thermal-hydraulics in the FA. In the past, validation studies to confirm applicability to thermal-hydraulics analysis in FAs have been performed by using measured data in water and sodium experiments. In this study, essential models in the codes to solve the important phenomena in the FA and necessary experiments for validation were listed systematically in order to assess the reliability of the codes, through developing an importance ranking table for the phenomena and a validation matrix according to the guide-line for the verification and validation (V&V). The ranking table was developed to decide the priority for validation. In addition, a validation matrix of experimental data and numerical models in the codes for the high priority phenomena in the ranking table were developed to confirm the sufficiency of the validation process. In future works, the validation study on large scale FA will be conducted for reliability improvement, and the codes will be validated with uncertainty quantification, by using this validation matrix.