Computer Codes

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, radionuclide transport, and materials performance 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
• Radiological Protection Computer Codes
• Materials Performance Codes

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 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 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.

Thermal-Hydraulics Codes

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.

• MACCS: The MELCOR Accident Consequence Code System (MACCS) code is the NRC code used to perform probabilistic offsite consequence assessments for hypothetical atmospheric releases of radionuclides.   The code models atmospheric transport and dispersion, emergency response and long-term protective actions, exposure pathways, early and long-term health effects, land contamination, and economic costs.  MACCS is used by U.S. nuclear power plant license renewal applicants to support the plant specific evaluation of severe accident mitigation alternatives (SAMA) as part of an applicant's environmental report for license renewal.  MACCS is also used in severe accident mitigation design alternatives (SAMDA) and severe accident consequence analyses for environmental impact statements for new reactor applications. The NRC uses MACCS in its cost-benefit assessments supporting regulatory analyses that evaluate potential new regulatory requirements for nuclear power plants.

• 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.

Radiological Protection Computer Code Analysis and Maintenance Program (RAMP) Codes

The NRC initiated the Radiation Protection Computer Code Analysis and Maintenance Program (RAMP) for the development, maintenance, improvement, and distribution of the NRC's collection of radiation protection, health physics, dose assessment, and emergency response computer codes. RAMP codes calculate dose for scenarios such as environmental assessment, nuclear power plant licensing, emergency response, atmospheric assessment, decommissioning, bioassay, and others.

RAMP provides a centralized management structure for code updates, distribution, modernization, applied research, training, and issue resolution. The RAMP user group of 3500+ active members include NRC staff, NRC and Agreement State licensees, private corporations, university/non-profit researchers, other U.S. federal agency staff, U.S. state/local officials, and national regulators across the world. Most RAMP codes are free for most users.

For the latest information on RAMP codes, user meetings, code access, and registration, visit the RAMP website.

RAMP's Premium Codes include:

• IMBA: RAMP's IMBA code (formerly IMBA Pro) is a suite of software modules for internal dosimetry that implements respiratory tract, GI-tract, tissue dosimetry, biokinetic and bioassay models as recommended by the International Commission on Radiological Protection (ICRP). Originally designed by the UK's Health Security Agency, the IMBA modules can estimate single or multiple intakes of different radionuclides and calculate resulting doses in the body and/or excrement for workers based on ICRP Publications 26/30 and 60/68, as well as U.S. regulations outlined in 10 CFR 835. IMBA provides a platform for conducting customized dose calculations with different user set parameters. The suite has functionalities such as performing simple and more complex dose calculations, vapor inhalation modelling, intake estimation for multiple regimes, bioassay quantities at different times from a specified intake, 740 radionuclides, and more.

• RASCAL: The Radiological Assessment System for Consequence Analysis computer code is used for making dose projections for atmospheric releases during radiological emergencies. RASCAL is used by the Protective Measures Team in the NRC's Operations Center for making independent dose and consequence projections during radiological incidents and emergencies. RASCAL was developed by NRC over 25 years ago to provide a tool for the rapid assessment of an incident or accident at an NRC-licensed facility and aid decision-making such as whether the public should evacuate or shelter in place. RASCAL evaluates atmospheric releases from nuclear power plants, spent fuel storage pools and casks, fuel cycle facilities, and radioactive material handling facilities. Its data is not the only criterion used by the local authorities during an accident, but certainly an important one.

• SNAP/RADTRAD: The Symbolic Nuclear Analysis Package/ RADionuclide, Transport, Removal, and Dose estimation code is a licensing analysis code used to show compliance with nuclear plant siting criteria for the site boundary radiation doses at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) and to assess the occupational radiation doses in the control room (CR) and /or Emergency Offsite Facility for various loss-of-coolant accidents (LOCA) and non-LOCA design basis accidents (DBAs). RADTRAD uses a combination of tables and numerical models of source term reduction phenomena to determine the time-dependent dose at the CR, EAB and LPZ for given DBA scenarios.

RAMP base codes:

• DandD: The DandD computer code is used by NRC licensees to demonstrate in an application for decommissioning a materials license (and the NRC to verify) that residual soil or building contamination at the licensed site following decontamination and decommissioning complies with the radiological dose criteria for license termination in 10 CFR 20, Subpart E. The computer code was designed to simplify decommissioning in cases where low levels of contamination exist.

• GALE: The gaseous and liquid effluent computer code, for pressurized-water and boiling-water reactors, which estimate the quantities of radioactivity released by a plant through liquid and atmospheric discharges during routine operations.

• GENII: GENII is a set of programs for estimating radionuclide concentrations in the environment and dose to humans from acute or chronic exposures from radiological releases to the environment or initial contamination conditions. It is part of a set of quality-assured and configuration-controlled safety analysis codes managed and maintained for the U.S. Department of Energy’s Safety Software Central Registry and the NRC.

• HABIT: The HABIT computer code is an integrated set of computer programs used mainly to estimate chemical exposures that personnel in the control room of a nuclear facility would be exposed to in the event of an accidental release of toxic chemicals.

• MILDOS: The MILDOS computer code calculates the radiological dose commitments received by individuals and the general population within an 80-km radius of an operating uranium recovery facility. In addition, air and ground concentrations of radionuclides are estimated for individual locations, as well as for a generalized population grid. Extra-regional population doses resulting from the transport of radon and export of agricultural produce are also estimated.

• NRCDose3: NRCDose3 is a software suite that integrates the functionality of three individual Fortran codes: LADTAP II, GASPAR II and XOQDOQ that were developed by the NRC and have been in use by NPP licensees and the NRC staff for assessments of routine liquid radioactive releases and offsite doses, routine gaseous radioactive effluents and offsite doses, and meteorological transport and dispersion, respectively. NRCDose3 is primarily used to support reactor licensing in the confirmatory evaluation on the environmental dose impacts of routine liquid and gaseous radiological effluent releases.

• NRC-RADTRAN: The NRC-RADTRAN computer code is used for risk and consequence analysis associated with routine, incident-free, transportation of radioactive materials and accidents that might occur during radioactive material transportation. NRC-RADTRAN will produce estimates of incident-free population dose, accident dose-risk, non-radiological traffic mortality, and a suite of individual dose estimates.

• PIMAL: The Phantom with Moving Arms and Legs is a graphical user interface that allows users to develop input decks for the Monte Carlo N-Particle (MCNP) code. PiMAL contains humanoid, equine, feline and canine phantom models which are considered an efficient and accurate tool for developing exposure models and performing dosimetry calculations for radiation workers and exposed members of the public.

• SIERRA ATD Module: The Software Integration for Environmental Radiological Release atmospheric transport and dispersion code used for analysis in siting, licensing, and environmental reviews for evaluating releases in cases of design-based accidents (from 100s of meters to 10 km), as well as normal effluent releases for sensitive receptors and populations up to 80 km. This module consolidates and modernizes the scientific functions of ARCON, PAVAN, and XOQDOQ into a single user interface and allows users to estimate relative concentrations based on hourly meteorological data for all three codes, rather than use of joint frequency distributions.

• Table Calculator: The Table Calculator is a user-friendly tool to facilitate a more comprehensive understanding of the calculations used to develop the low-level radioactive waste classification tables. The tool runs in the GoldSim Player, and it allows users to trace the original calculations and observe the effects of changes in parameter values by running the original calculations with the original or updated data.

• V+: The V+ (formerly VARSKIN+) computer code is a comprehensive radiation dose assessment tool designed to evaluate occupational and medical exposure scenarios. The software includes six specialized modules, each tailored to a specific type of radiation dose calculation: SkinDose, WoundDose, NeutronDose, EyeDose, Radiological Toolbox, and ExtravDose. V+ is RAMP’s most popular code.

RAMP joint-sponsored codes:

• RESRAD: The Residual Radioactivity family of codes analyze potential human and biota radiation exposures from the environmental contamination of residual radioactive materials. The codes use pathway analysis to evaluate radiation exposure and associated risks, and to derive cleanup criteria or authorized limits for radionuclide concentrations in the contaminated source medium. This code is jointly sponsored with and distributed by Argonne National Laboratory.

• VSP: The Visual Sample Plan computer code is a tool that helps ensure the right type, quality, and quantity of data are gathered to support confident decisions and provides statistical evaluations of the data with decision recommendations. VSP couples' site, building, and sample location visualization capabilities with optimal sampling design and statistical analysis strategies. This code is jointly sponsored with and distributed by Pacific Northwest National Laboratory.

RAMP-affiliated codes:

• CAP88: CAP88-PC (Clean Air Act Assessment Package - 1988) is a computer code for estimating the dose and risk from emissions of radioactive material to the air. Version 4.1.1 is the most current CAP88-PC and is a regulatory compliance tool under the National Emissions Standard for Hazardous Air Pollutants (NESHAPs), Subpart H. The CAP88–PC program is a well-established and validated code for the purpose of making comprehensive dose and risk assessments. The Gaussian plume model used in CAP88–PC to estimate dispersion in air is one of the most used models for dispersion modeling. CAP 88-PC estimates the average dispersion of radionuclides released from up to six sources, that may be either elevated stacks, such as smokestack, or uniform area sources, such as pile of uranium mill tailings. It produces results that agree with experimental data as well as any model, is easy to work with, and is consistent with the random nature of turbulence. This code is distributed by the U.S. Environmental Protection Agency.

• Turbo FRMAC: Turbo FRMAC code performs complex calculations to quickly evaluate radiological hazards during an emergency response by assessing impacts on the public, workers, and the food supply. Turbo FRMAC can be used to evaluate the hazard from a wide variety of radiological incidents, such as a nuclear power plant emergency. Turbo FRMAC calculations are based on methods established by the Federal Radiological Monitoring and Assessment Center (FRMAC). This code is distributed by Sandia National Laboratories.

Materials Performance Codes

Materials performance codes are used to evaluate proactive approaches for the management of aging degradation mechanisms and structural integrity issues affecting nuclear power plant components:

• ASME Section III, Division 5 Design Tool

  This tool, consisting of scripts written in the Python code, an open source computer language,  executes the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section III, Division 5 (Section III-5) design rules for high temperature metallic components.  The tool was developed under a contract with Argonne National Laboratory (ANL).  The tool facilitates checking a design for an advance reactor component against all the Section III-5, Subsection HB, Subpart B (HBB) design criteria for primary load limits, strain limits, and creep-fatigue damage.  Applied stresses determined using any commercial finite element analysis software can be entered in an Excel spreadsheet, which is then read by the tools and checked against the various Section III, Division 5 design criteria.  The tool implements the automatic design evaluation for the Section III-5 rules for elastic analysis and the Code Case N-861 and N-862 rules for elastic-perfectly plastic (EPP) analysis, but does not implement the Section III-5 rules for design by inelastic analysis.

• LEAPOR is a computer code used to calculate the rate of leakage of water flowing from a through-wall crack in a pipe, an essential part of a leak-before-break analysis. LEAPOR serves as the leak rate module included in the xLPR code. A stand-alone version of LEAPOR includes a graphical user interface to provide functionality outside of xLPR.

• xLPR: Extremely Low Probability of Rupture. A state-of-the-art probabilistic fracture mechanics code for piping applications. The code models failure probabilities associated with nuclear power plant piping system components subject to active degradation mechanisms. Its core capabilities include modeling fatigue, stress-corrosion cracking, inservice inspection, chemical and mechanical mitigation, leakage rates, and seismic effects.

  xLPR was jointly developed by the NRC's Office of Nuclear Regulatory Research and the Electric Power Research Institute. Code development was a multi-year effort that built on the results of a successful pilot study. The code was designed, built, and tested under a rigorous software quality assurance program and provides regulators, industry, researchers, and the public with new quantitative capabilities to analyze the risks associated with nuclear power plant piping systems subject to active degradation mechanisms.

• FAVPRO: "Fracture Analysis of Vessels – Probabilistic" is a probabilistic fracture mechanics code for reactor pressure vessel integrity analysis. The code models large Light-Water Reactor (LWR) vessels such as those employed in Gen II and Gen III Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). FAVPRO can perform deterministic and probabilistic assessments of reactor pressure vessel fracture for any number of transients in one execution of FAVPRO. A wide variety of analysis options are available to analysts to model different physical phenomena (weld residual stresses, warm pre-stress, crack propagation by cleavage fracture or stable and unstable ductile tearing, variable failure criteria, predictions of embrittlement, etc.)

  FAVPRO was developed by the NRC, along with commercial contractors NUMARK Inc. and Archaeologic Inc., to replace the legacy FAVOR code suite. FAVPRO was released in 2024 and integrates the separate FAVOR modules into one unified executable. The legacy FAVOR source code was refactored using modern object-oriented parallel Fortran 2018 standards while following state-of-practice Software Quality Assurance and Verification and Validation standards. The new FAVPRO code allows for parallel execution of probabilistic analyses, providing a major performance improvement over FAVOR. Input and output files in FAVPRO have also been updated to the modern JavaScript Object Notation format. An Excel-based automated input generator (FAVPRO-AIG) supports the creation of FAVPRO input files and a Python-based data visualization tool (FAVPRO-VT) supports the interpretation of output files.