Overview

Activities in Reactor Transient and Safety Analysis research and development fulfill a primary goal of the Nuclear  Engineering (NE) Division to promote improvements in safe and reliable operation of present and future nuclear reactor systems. These activities range from development of new models and modeling techniques to analysis applications supporting design, construction, and operation of nuclear reactor facilities.

Work in the Reactor Transient and Safety Analysis area is closely coupled to other major capability areas in the NE Division. The Reactor Physics and Fuel Cycle Analysis capability in NE provides definition of reactor physics and design information for analysis of reactor design basis and accident analyses. The Reactor Safety Experimentation area provides experimental and test data for validation of analysis models. The Heat Transfer and Fluid Mechanics and Engineering and Structural Mechanics areas provide benchmark analysis models and computational results for reactor analysis model verification. And the Advanced Computation and Visualization area provides new model solution techniques as well as new ways of displaying computational results. Information produced by Reactor Transient and Safety Analysis directly supports new design development, safety verification of existing reactors, applications to regulatory authorities for operation of reactor facilities, and safety assessments including those produced in the Risk Methodology and Evaluation (PRA and PSA) area.

Activities in the Reactor Transient and Safety Analysis area are conducted in support of several NE programs and projects, including the Advanced Reactor Development and Technology Program, the Advanced Fuel Cycle Program, and the Fissile Material Disposition Program.

Analyses performed in this area utilize large-scale, integrated Computer Codes designed to model entire reactors and their associated engineered systems. The overall reactor systems model consists of many coupled individual models of heat transfer, fluid dynamics, reactor neutron kinetics, and reactor structural mechanics. These phenomenological models combine to provide a transient simulation of the behavior of all of the major components in a nuclear reactor plant, including the fuel and coolant, the coolant systems and components (vessels, pipes, valves, pumps, and heat exchangers), and the plant control and protection systems. Work at Argonne in the Reactor Transient and Safety Analysis area has produced the SAS4A and SASSYS-1 computer codes for safety and systems analysis of liquid-metal-cooled nuclear reactors, and the SAS-DIF3DK computer code for coupled spatial kinetics/thermal-hydraulics analysis of water-cooled nuclear reactors. In addition to these Argonne codes, transient and safety analysis computer codes developed at other US and international scientific institutes and research organizations are maintained for use in Reactor Transient and Safety Analysis in the NE Division.