In this paper, development and preliminary validation of an in-house code LETHAC (lead-based fast reactor thermal-hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal-hydraulic transient performance of a lead-bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most-used LBE (lead-bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN-II 61-rod wire-wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire-wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead-bismuth cooled systems.