|
|
Home > About NRC
> How We Regulate > Research Activities
> Computer Codes
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, 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:
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 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 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.
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. 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.
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 the Radiation Safety Information Computational Center (RSICC)
 . For additional information, see NUREG/CR-6918, "VARSKIN 3: A Computer Code
for Assessing Skin Dose from Skin Contamination."
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 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.)
|
