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:
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.
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 Radiological Protection Computer Code Analysis and Maintenance Program (RAMP) for the development, maintenance, and distribution of the NRC's vast array of radiation protection, dose assessment, and emergency response computer codes. The benefits of RAMP include: access to the most current versions of the code; code maintenance, development, benchmarking, and uncertainty studies; a cooperative forum to resolve code errors and inefficiencies; technical basis documents and user guidelines for applying the codes, and periodic meetings to share experiences, discuss code development; and, periodic training on the codes. For information regarding RAMP codes or how to join, please visit the RAMP website.
- 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.
- 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.
VARSKIN: Computer code for calculating Skin dose. VARSKIN assesses compliance with the dose criteria of 10 CFR Part 20. The code is used to perform confirmatory calculations of licensees' submittals regarding skin dose (from both beta and gamma sources) estimates at any skin depth or skin volume, with point, disk, cylindrical, spherical, or slab (rectangular) sources, and even enables users to compute doses from multiple sources.
Five different predefined source configurations are available in VARSKIN that allow simulations of point, disk, cylinder, sphere, and slab sources on the skin. Recent improvements to earlier VARSKIN versions include an enhanced photon dosimetry model, as well as models to account for air gap and cover materials for photon dosimetry. Additionally, VARSKIN has been updated to better predict beta dosimetry in shallow skin targets. Although the user can choose any dose-averaging area, the default area for skin dose calculations in VARSKIN is 10 square centimeters, to conform to regulatory requirements pursuant to Title 10 of the Code of Federal Regulations, Section 20.1201(c). Data entry is condensed to a single screen, a variety of unit options are provided (including both British and International System (SI) units), and the source strength can be entered in units of total activity or distributed in units of activity per unit area or activity per unit volume. The output page and the user's ability to add radionuclides to the library are greatly simplified. A library file contains data on gamma rays, X rays, beta particles, internal conversion electrons, and Auger electrons. VARSKIN allows the user to build a customized library of exposure radionuclides. Finally, an extensive, context-sensitive help file is made available for VARSKIN to provide guidance and to offer new users a tutorial in the use of the skin dosimetry software.
The enhanced photon model accounts for photon attenuation, charged particle buildup, and electron scatter at all depths in skin. The model allows for volumetric sources and clothing/air gaps between source and skin. The beta dosimetry model has been upgraded to better account for beta energy loss and particle scatter. Dose point kernels are now Monte Carlo based and the code agrees very well with the EGSnrc Monte Carlo code.
Radiological Toolbox: The NRC developed the radiological toolbox as a means to quickly access databases needed for radiation protection, shielding, and dosimetry calculations. The toolbox is essentially an electronic handbook with limited computational capabilities beyond those of unit conversion. Further revisions of the toolbox are planned as the need for additional data is identified by NRC staff and other users. The toolbox contains radioactive decay data, biokinetic data, internal and external dose coefficients, elemental composition of a large number of materials, radiation interaction coefficients, kerma coefficients, and other tabular data of interest to the health physicist, radiological engineer, and others working in fields involving radiation. The toolbox includes a means to export the tabular data to an Excel worksheet for use in further calculations. It operates in a Windows environment.
HABIT: Computer code for evaluating control room HABIT ability. The HABIT 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.
GALE: The FORTRAN based gaseous and liquid effluent (GALE) code estimates the quantities of radioactivity released by a plant through liquid and atmospheric discharges during routine operations for pressurized-water reactors (PWR) and boiling-water reactors (BWR).
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.
Radionuclide Transport Codes (for License Termination and Decommissioning)
Radionuclide transport and decommissioning codes provide dose analyses in support of license termination and decommissioning:
- 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.)
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.
FAVOR: the "Fracture Analysis of Vessels – Oak Ridge" is a probabilistic fracture mechanics code for reactor pressure vessel integrity analysis. The code is validated for modeling large Light-Water Reactor (LWR) vessels such as those employed in Gen II and Gen III Pressurized Water (PWR) Reactors and Boiling Water Reactors (BWR). FAVOR can statistically model the variations in vessel properties and the uncertainty in flaw populations to produce holistic vessel integrity predictions for any number of transients whose frequency distributions can be specified. A wide array of options are available to the analyst for the different physical phenomena modeled in the code (warm pre-stress, crack propagation and arrest, ductile tearing failure, material embrittlement, etc.).
FAVOR was developed by Oak Ridge National Laboratory for the NRC in the 1990's, and is now maintained and developed in-house by NRC with the help of commercial contractors. As of 2021, FAVOR is being refactored and modernized to meet modern state-of-practice Software Quality Assurance and Verification and Validation standards. This includes implementing agile software development practices and updating the source code to modern object-oriented parallel Fortran 2018 standards.
Page Last Reviewed/Updated Wednesday, February 10, 2021