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Computer Codes

NRC 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 Codes
• Reactor Kinetics
• Thermal Hydraulics
• Severe Accident Codes
• Radionuclide Transport and Decommissioning Codes
• Design Basis Accident (DBA) Codes

Fuel Behavior Codes

Fuel 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 Kinetics

Reactor 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 Hydraulics

Advanced 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:

• TRACE: The TRAC/RELAP Advanced Computational Engine. A modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety codes - TRAC-P, TRAC-B and RELAP.   It is able to analyze large/small break LOCAs and system transients in both PWRs and BWRs. The capability exists to model thermal hydraulic phenomena in both 1-D and 3-D space. This is NRC's flagship thermal-hydraulics analysis tool.
• SNAP: Symbolic Nuclear Analysis Package is a graphical user interface with pre-processor and post-processor capabilities that assists code users in the development of TRACE and RELAP5 input decks and in running the codes.
• RELAP5: Reactor Excursion and Leak Analysis Program - Small break LOCA and system transient analysis tool for PWRs or BWRs. It has the capability to model thermal hydraulic phenomena in 1-D volumes. While this code still enjoys widespread use in the nuclear community, active maintenance will be phased out in the next few years as usage of TRACE grows.
• Other legacy tools that are no longer actively supported include
  • TRAC-P: Large break LOCA and system transient analysis tool for PWRs. Capability to model thermal hydraulic phenomena in 1-D or 3-D components.
  • TRAC-B: Large and small break LOCA and system transient analysis tool for BWRs. Capability to model thermal hydraulic phenomena in 1-D or 3-D components.
  • CONTAIN: Containment transient analysis tool for PWRs or BWRs. Capability to model thermal hydraulic phenomena (within a lumped-parameter framework) for existing containment designs.

Severe Accident Codes

Severe accident codes are used to model the progression of accidents in light water reactor nuclear power plants.

• MELCOR: Integral Severe Accident Analysis Code: Fast-Running, parametric models.
• SCDAP/RELAP5: Integral Severe Accident Analysis Code: Uses detailed mechanistic models.
• CONTAIN: Integral Containment Analysis Code: uses detailed mechanistic models. (CONTAIN severe accident model development was terminated in the mid-1990s.) The MELCOR code has similar containment capabilities (but less detailed in some areas) and should generally be used instead of CONTAIN.
• IFCI: Integral Fuel-Coolant Interactions Code.
• VICTORIA: Radionuclide Transport and Decommissioning Codes: Radionuclide transport and decommissioning codes provide dose analyses in support of license termination and decommissioning.

Radionuclide Transport and Decommissioning Codes

Radionuclide transport and decommissioning codes provide dose analyses in support of license termination and decommissioning.

• DandD: A code for screening analyses for license termination and decommissioning. The DandD software automates the definition and development of the scenarios, exposure pathways, models, mathematical formulations, assumptions, and justifications of parameter selections documented in Volumes 1 and 3 of NUREG/CR-5512.
• Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Codes: The existing deterministic RESRAD 6.0 and RESRAD-BUILD 3.0 codes for site-specific modeling applications were adapted by Argonne National Laboratory (ANL) for NRC regulatory applications for probabilistic dose analysis to demonstrate compliance with the NRC's license termination rule (10 CFR Part 20, Subpart E) according to the guidance developed for the Standard Review Plan (SRP) for Decommissioning. (The deterministic RESRAD and RESRAD-BUILD codes are part of the family of codes developed by the U.S. Department of Energy. The RESRAD code applies to the cleanup of sites and the RESRAD-BUILD code applies to the cleanup of buildings and structures.)

Design Basis Accident (DBA) Codes

DBA codes are used to determine the time-dependent dose at specified location for given accident scenarios.

• RADTRAD: A simplified model for RADionuclide Transport and Removal And Dose Estimation. The RADTRAD code uses a combination of tables and numerical models of source term reduction phenomena to determine the time-dependent dose at specified locations for a given accident scenario. It also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The RADTRAD code can be used to assess occupational radiation exposures, typically in the control room; to estimate site boundary doses; and to estimate dose attenuation due to modification of a facility or accident sequence.

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Tuesday, March 13, 2007