Both Alloy 617 and Alloy 230 have been considered the most promising structural materials for the Very High Temperature Reactor (VHTR). In this study, mechanical properties of both alloys were examined by performing tensile tests at three different strain rates and at temperatures up to 1000°C. This range covers time-dependent (plasticity) to time-independent (creep) deformations. Strain-rate sensitivity analysis for each alloy was conducted in order to approximate the long-term flow stresses. The strain-rate sensitivities for the 0.2% flow stress were found to be temperature independent (m ≈ 0) at temperatures ranging from room temperature to 700°C due to dynamic strain aging. At elevated temperatures (800-1000°C), the strain-rate sensitivity significantly increased (m > 0.1). Compared to Alloy 617, Alloy 230 displayed higher strain-rate sensitivities at high temperatures. This leads to a lower estimated long-term flow stresses. Results of this analysis were used to evaluate current American Society of Mechanical Engineers (ASME) allowable design limits. According to the comparison with the estimated flow stresses, the allowable design stresses in ASME B&PV Code for either alloy did not provide adequate degradation estimation for the possible long-term service life in VHTR. However, rupture stresses for Alloy 617, developed in ASME code case N-47-28, can generally satisfy the safety margin estimated in the study following the strain-rate sensitivity analysis. Nevertheless, additional material development studies might be required, since the design parameters for rupture stresses are constrained such that current VHTR conceptual designs cannot satisfy the limits.