Large thermal-hydraulic systems computer codes are most often applied to investigate safety issues in existing nuclear facilities. One such code is applied to aid the design process for a proposed state-of-the-art research reactor. The RELAP5 computer code is used to simulate system response to hypothetical loss-of coolant accidents (LOCAs) in an early design of the Advanced Neutron Source (ANS). Among accident scenarios, a LOCA event is expected to be one of the most challenging to the ANS reactor core; similar analyses for other accident types are in progress. This is the first detailed study of ANS transient system response during accidents, and the outcome of the analysis is used to benefit the design process. The ANS model used is based on an early (preconceptual) cool ing system design layout. This early design has since been superseded by an improved design that is partly based on the results of these studies. The calculated responses of the early design to representative LOCA events are described; the simulations indicate that fuel melting and damage would be experienced for medium and large breaks. The effectiveness of employing a gascharged accumulator on the primary coolant system for preventing fuel damage following medium- and large-break LOCAs is evaluated. As a result of this evaluation, the new ANS design incorporates such accumulators. Analysis uncertainties are addressed, and the findings from this study that were used for the next phase of ANS design are highlighted.