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Computer CodesNRC uses computer codes to evaluate thermal-hydraulic conditions, fuel behavior, and reactor kinetics during various operating and postulated accident conditions. Results from applying the codes support decisionmaking for risk-informed activities, the review of licensees' codes and performance of audit calculations, and the resolution of other technical issues. Code development is directed toward improving the realism and reliability of code results and making the codes easier to use. On this page:
Fuel Behavior CodesFuel behavior codes are used to evaluate fuel behavior under various reactor operating conditions. FRAPCON-3 is a computer code used for steady-state and mild transient analysis of the behavior of a single fuel rod under near-normal reactor operating conditions. FRAPTRAN is a computer code used for transient and design basis accident analysis of the behavior of a single fuel rod under off-normal reactor operation conditions. Reactor KineticsReactor kinetics are used to obtain reactor transient neutron flux distributions. PARCS: The Purdue Advanced Reactor Core Simulator (PARCS) is a computer code that solves the time-dependent two-group neutron diffusion equation in three-dimensional Cartesian geometry using nodal methods to obtain the transient neutron flux distribution. The code may be used in the analysis of reactivity-initiated accidents in light water reactors where spatial effects may be important. It may be run in the stand-alone mode or coupled to other NRC thermal-hydraulic codes such as RELAP5. Thermal HydraulicsAdvanced computing plays a critical role in the design, licensing and operation of nuclear power plants. The modern nuclear reactor system operates at a level of sophistication whereby human reasoning and simple theoretical models are simply not capable of bringing to light full understanding of a system's response to some proposed perturbation, and yet, there is an inherent need to acquire such understanding. Over the last 30 years or so, there has been a concerted effort on the part of the power utilities, the U. S. Nuclear Regulatory Commission (USNRC), and foreign organizations to develop advanced computational tools for simulating reactor system thermal-hydraulic behavior during real and hypothetical transient scenarios. In particular, thermal hydraulics codes are used to analyze loss of coolant accidents (LOCAs) and system transients in light water nuclear reactors. The lessons learned from simulations carried out with these tools help form the basis for decisions made concerning plant design, operation, and safety. The United States Nuclear Regulatory Commission (NRC) and countries in the international nuclear community have agreed to exchange technical information on thermal-hydraulic safety issues related to reactor and plant systems. Under the terms of their agreements, the NRC provides these member countries the latest versions of its thermal-hydraulic systems analysis computer codes to help evaluate the safety of planned or operating plants in each member's country. To help ensure these analysis tools are of the highest quality and can be used with confidence, the international partners perform and document assessments of the codes for a wide range of applications, including identification of code improvements and error corrections. The computer codes developed by the NRC include the following:
Severe Accident CodesSevere accident codes are used to model the progression of accidents in light water reactor nuclear power plants.
Radionuclide Transport and Decommissioning CodesRadionuclide transport and decommissioning codes provide dose analyses in support of license termination and decommissioning.
Design Basis Accident (DBA) CodesDBA codes are used to determine the time-dependent dose at specified location for given accident scenarios.
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