Abstract: In this study, the activation behaviour and shielding design of spent fuel assembly components were investigated, focusing specifically on the post-shutdown storage of guide tubes and in-core detectors. Activation analysis and dose assessment of reactor components are essential for decommissioning planning, cost and schedule estimation, as well as for the preparation of radiation protection and emergency response measures. At present, existing domestic computational tools are constrained by incomplete nuclide databases and inadequate simulation of irradiation histories, which compromises the reliability and cost-effectiveness of safety analyses for reprocessing. To address these limitations, a coupled computational framework by integrating a Monte Carlo neutron transport code with a depletion/activation code was established in this paper. The framework adopted an iterative time-stepping scheme to update the nuclide inventory, neutron flux, and group-wise effective reaction cross-sections, and enabled continuous burnup-dependent activation analysis. First, the proposed method was validated against the OECD/NEA fast reactor burn-up benchmark. The calculated isotopic evolution agreeds well with the benchmark results, indicating acceptable predictive capability of the coupled transport-depletion approach. Subsequently, a CAP1000 17×17 fuel assembly scenario was modelled, including the fuel management and multi-cycle irradiation history. After multiple irradiation cycles and post-shutdown cooling, the guide tube material (M5 alloy) was analyzed. Next, activation calculations were performed. Dose rates at various distances from the detector and guide tube surfaces were calculated for different cooling time. To validate the detector-related dose, long-term RIC (¹⁰³Rh) measurement data were compared with simulation results under steady-state conditions. Finally, a preliminary shielding scheme was proposed. Based on the calculated gamma source term, an initial shielding design was developed for the detector storage configuration. A silicon-polymer/tungsten composite material was evaluated using dose attenuation metrics. The formulation with a high tungsten content significantly improves gamma shielding performance while retaining practical advantages such as flexibility and processability. Thus, this study provides a practical workflow that links activation analysis, dose assessment and preliminary shielding optimisation for spent-fuel assembly detector management. Current limitations include the absence of impurity-driven pathways (notably ⁶⁰Co), incomplete coupling with neutron shielding, and a lack of fabrication and testing of the new composite material. These are identified as priorities for future work.