The U.S. Nuclear Regulatory Commission (NRC) uses computer codes to model and evaluate fuel behavior, reactor kinetics, thermal-hydraulic conditions, severe accident progression, time-dependent dose for design-basis accidents, emergency preparedness and response, health effects, and radionuclide transport, during various operating and postulated accident conditions. Results from applying the codes support decisionmaking for risk-informed activities, review of licensees' codes and performance of audit calculations, and 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. For more information, see the following code categories on this page:
- Probabilistic Risk Assessment Codes
- Fuel Behavior Codes
- Reactor Kinetic Codes
- Thermal-Hydraulics Codes
- Severe Accident Codes
- Design-Basis Accident (DBA) Codes
- Emergency Preparedness & Response (EPR) Codes
- Health Effects/Dose Calculation Codes
- Radionuclide Transport Codes (for License Termination/Decommissioning)
Further information about obtaining these computer codes can be found by following the "Obtaining the Codes" link (also on the left side of this webpage).
Probabilistic Risk Assessment Codes
- SAPHIRE: Systems Analysis Programs for Hands-on Integrated Reliability (SAPHIRE) is used for performing probabilistic risk assessments.
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
- 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 Codes
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.
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 NRC, 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 NRC and other 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 thermal-hydraulics 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 pressurized- and boiling-water reactors (PWRs and BWRs). The capability exists to model thermal hydraulic phenomena in both one-dimensional (1-D) and three-dimensional (3-D) space. This is the NRC's flagship thermal-hydraulics analysis tool.
- SNAP: The Symbolic Nuclear Analysis Package is a graphical user interface with pre-processor and post-processor capabilities, which assists users in developing TRACE and RELAP5 input decks and running the codes.
- RELAP5: The Reactor Excursion and Leak Analysis Program is a tool for
analyzing small-break LOCAs and system transients in 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.
- Legacy tools that are no longer actively supported include
the following thermal-hydraulics codes:
- 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.
- MACCS2: Accident Consequence Analysis Code: The computer code used to calculate dispersion of radioactive material to the environment and the population. The MACCS2 code uses a dose-response model to determine the health consequences of a severe accident in terms of early fatalities (how many people in a population would die in the weeks or months following exposure) and latent cancer risk (how many people in a population would contract a fatal cancer as a result of exposure). MACCS2 originated as an acronym for the MELCOR Accident Consequence Code System, but is now commonly known simply as the MACCS2 Accident Consequence Analysis Code.
- 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
- VICTORIA: Radionuclide Transport and Decommissioning Codes: Radionuclide transport and decommissioning codes provide dose analyses in support of license termination and decommissioning.
Design-Basis Accident (DBA) Codes
DBA codes are used to determine the time-dependent dose at a specified location for a given accident scenario:
- 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. 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. RADTRAD 3.03 is available from the Radiation Safety Information Computational Center (RSICC)
Emergency Preparedness & Response (EPR) Codes
EPR codes compute power reactor source terms, airborne transport of activity, and the resulting doses to allow easy comparison to EPA protective action guidelines:
RASCAL: Radiological Assessment Systems for Consequence AnaLysis. The RASCAL code evaluates releases from nuclear power plants, spent fuel storage pools and casks, fuel cycle facilities, and radioactive material handling facilities and is designed for use by the NRC in the independent assessment of dose projections during response to radiological emergencies. Obtain the latest information on RASCAL including version 4.3. There is no cost associated with receipt of this code.
Health Effects/Dose Calculation Codes
Health effects/dose calculation codes are used to model and assess the health implications of radioactive exposure and contamination.
- VARSKIN:The NRC sponsored the development of the
VARSKIN code in the 1980s, to assist licensees in
demonstrating compliance with Paragraph (c) of Title 10,
Section 20.1201, of the Code of Federal Regulations (10 CFR 20.1201),
"Occupational Dose Limits for Adults." Specifically,
10 CFR 20.1201(c) requires licensees to have an approved
radiation protection program that includes established protocols
for calculating and documenting the dose attributable to
radioactive contamination of the skin. Since that time, the
code has been significantly enhanced to simplify data entry
and increase efficiency. VARSKIN 3 is available from
Safety Information Computational Center (RSICC) . For additional information,
"VARSKIN 3: A Computer Code for Assessing Skin Dose from
Since the release of VARSKIN 3 in 2004, the NRC staff has compared its dose calculations for various energies and at various skin depths, with doses calculated by the Monte Carlo N-Particle Transport Code System (MCNP ) developed by Los Alamos National Laboratory (LANL ). That comparison indicated that VARSKIN 3 overestimates the dose with increasing photon energy. For that reason, the NRC is sponsoring a further enhancement to replace the existing photon dose algorithm, develop a quality assurance program for the beta dose model, and correct technical issues reported by users. To facilitate that enhancement, we encourage you to Contact Us, if you are aware of any problems or errors associated with the VARSKIN code.
Radionuclide Transport Codes (for License Termination and Decommissioning)
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.)