United States Nuclear Regulatory Commission - Protecting People and the Environment

Reviewing Accident Tolerant Fuel: In-Reactor Performance | Fuel Cycle, Transportation, and Storage | Probabilistic Risk Assessment | Independent Confirmatory Calculation

Independent Confirmatory Calculations

On this page

This page includes links to files in non-HTML format. See Plugins, Viewers, and Other Tools for more information.


Independent confirmatory calculations are one of the tools that the NRC may use in its safety reviews. Confirmatory calculations performed with computer codes provide insight on the fuel and reactor systems behavior and potential consequences of transient and accident scenarios. In addition, sensitivity studies help to identify risk significant contributors to the safety analyses and assist in focusing the staff’s review.

Computer codes used for ATF modeling capabilities will likely be based on smaller data sets than those of the current fuel, resulting in greater uncertainty in the safety analyses’ results. The larger uncertainty could impact allowable margins for reactor operations or material transport. For these reasons, confirmatory calculation capabilities will be critical for generating confidence in the safety review of ATF.

There are four technical disciplines needing calculation capability development to support NRC safety reviews of ATF: (1) fuel performance, (2) thermal hydraulics, (3) neutronics, and (4) severe accidents. The NRC has developed a suite of computer codes to analyze these disciplines, and they have been used successfully to support regulatory decision-making for decades. Continued development of these codes ensures that the NRC has the capability to analyze ATF designs.

To top of page

Fuel Performance

Fuel performance codes calculate the temperature, pressure, and deformation of nuclear fuel and cladding during long-term in-reactor burnup and transient conditions. The NRC staff uses fuel performance codes during licensing reviews to see whether specified acceptable fuel design limits are maintained and to provide initial conditions to thermal hydraulics codes for design-basis accident (DBA) analysis. Additionally, the NRC uses fuel performance codes to support the safety limits for loading and storing spent nuclear fuel in dry casks.

The U.S. NRC's longstanding fuel performance codes, FRAPCON and FRAPTRAN, have been merged into a single, modern code called "FAST" (Fuel Analysis under Steady-state and Transients). Recent improvements to FAST include the creation of a material properties library that allows code developers to easily update properties for existing materials or to add properties for new materials for ATF. FAST currently includes properties for some ATF concepts – included Cr-coated cladding and iron-chromium-aluminum (FeCrAl) cladding – based on information available in the public literature. The NRC is working with the U.S. Department of Energy, Electric Power Research Institute, and the fuel vendors to obtain additional material property data, as well as the integral data needed to validate the codes.

To top of page

Thermal Hydraulics

Thermal hydraulics is the physics of fluid flow and energy (e.g., heat) transfer, as well as the interactions between fluid flow, energy, and the surrounding structures. Thermal hydraulics computer codes are used to analyze loss of coolant accidents (LOCAs) and system transients in light-water nuclear reactors. Results from these codes are used to verify the licensees are in compliance with applicable regulatory limits, such as the 2200° Fahrenheit peak cladding temperature limit stated in the emergency core cooling success criteria in 10 CFR 50.46.

The TRAC/RELAP Advanced Computational Engine (called TRACE) is the latest in a series of advanced, best-estimate reactor systems computer codes developed by the NRC. TRACE has been designed to perform best-estimate analyses of LOCAs, operational transients, and other accident scenarios in pressurized light-water reactors (PWRs) and boiling light-water reactors (BWRs). It may also model phenomena occurring in experimental facilities designed to simulate transients in reactor systems. TRACE includes simplified fuel performance models for current fuel designs. More recently, TRACE has been coupled to FAST (described above) and the U.S. Department of Energy's BISON code, which provides more detailed information about fuel performance during transients. This coupling also allows TRACE to benefit from new material properties added to FAST and BISON.


Neutronics is the study of the motions and interactions of neutrons with materials. Neutronics calculations are an integral part of the confirmatory review process for fuel changes because they provide decay heat rates, core power, and reactivity values used by thermal-hydraulic and fuel performance codes. Neutronics analysis is also needed to determine the mass of radioactive particles in the fuel for severe-accident/consequence analyses that are required by various regulations. Finally, neutronics analysis is performed to support decisions for spent fuel pool heat loading and amount of fission products in the fuel after it is removed from the reactor. See Task 4.c of the ATF Project Plan (pages B-8 through B-10) for more examples of the role of neutronics analyses in NRC reviews.

The NRC's main neutronics codes are SCALE and PARCS. The NRC staff uses SCALE to support licensing reviews in the areas of criticality safety and shielding. SCALE is also used to generate decay heat and radioactive material masses used in fuel performance, thermal hydraulic, and severe accident and source term calculations, and to determine simplified neutronics parameters used in the PARCS code. PARCS is used to calculate the reactor power as a function of time and position in the core. It has been incorporated in NRC’s TRACE thermal hydraulics code to provide reactor power during an accident scenario.

Severe Accidents and Source Term

Regulatory source terms are the types and amounts of radioactive or hazardous material released following an accident. The use of source terms is deeply embedded in the NRC's regulatory policy and practices as the current licensing process has evolved over the past 50 years. Regulatory source terms developed using severe accident analysis tools are used to verify that the plant design meets the safety and numerical radiological criteria set forth in the NRC’s regulations. Severe accident analysis is also used to support development of probabilistic risk assessment success criteria; to inform emergency preparedness and incident response procedures; and to assess the environmental impact of an accident.

The severe accident code used by the NRC is called MELCOR. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications.

The MELCOR code is well suited for developing a regulatory source term with flexible models to support the evaluation of differences between current standard fuel and ATF concepts. The MELCOR models are general in nature and may be adjusted to reflect differing properties of advanced fuels. Models for iron-chromium-aluminum (FeCrAl) cladding have recently been incorporated in MELCOR and are being validated using experimental data from the QUENCH international program led by Karlsruhe Institute of Technology in Germany. The upcoming severe accident PIRT exercise is expected to provide additional information that may support MELCOR development for ATF.

The various regulatory source terms are used in conjunction with DBA assumptions to confirm that the licensee meets the applicable radiological criteria contained within the NRC's regulations. When existing radiological regulatory requirements, guidance, and procedures were established, ATF, higher burnup, and increased enrichment fuel designs were not incorporated because they were not under development at that time. Therefore, to assess the radiological consequence of these new fuel designs, the NRC utilizes a computer code called Symbolic Nuclear Analysis Package/RADionuclide Transport, Removal and Dose Estimation (SNAP/RADTRAD). The NRC uses the output from MELCOR as an input to SNAP/RADTRAD to perform the radiological consequences analyses. These analyses assist the NRC to determine if the new fuel designs demonstrate compliance with the applicable radiological criteria in the regulations.

To top of page

Page Last Reviewed/Updated Monday, August 03, 2020