SCALE Overview


The SCALE code system is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of the Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic solvers and three Monte Carlo radiation transport solvers selected based on the user’s desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay
calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s GUIs assist with accurate system modeling and convenient access to desired results.


The history of the SCALE code system dates to 1969, when ORNL began providing the transportation package certification staff at the US Atomic Energy Commission (AEC) with computational support in the use of the new KENO code. KENO was used to perform criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976, the AEC certification staff relied on ORNL personnel to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analysis of transportation packages. However, the certification staff learned that users had difficulty in becoming proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the certification staff was moved to the US Nuclear Regulatory Commission (NRC), the NRC proposed development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of SCALE as a comprehensive modeling and simulation suite for nuclear safety analysis and design was born.

The NRC staff provided ORNL with some general development criteria for SCALE: (1) focus on applications related to nuclear fuel facilities and package designs, (2) use well-established computer codes and data libraries, (3) design an input format for the occasional or novice user, (4) prepare standard analysis sequences (control modules) to automate the use of multiple codes (functional modules) and data to perform a system analysis, and (5) provide complete documentation and public availability. With these criteria, the ORNL staff laid out the framework for the SCALE system and began development efforts. The initial version of SCALE (Version 0) was distributed in July 1980. Although the capabilities of the system continue to evolve, the philosophy established with the initial release still serves as the foundation of this year’s SCALE 6.2 release, more than 35 years later.

Capabilities of SCALE

A primary goal of SCALE is to provide robust calculations while reducing requirements for user input and knowledge of the intricacies of the underlying code and data architecture. SCALE provides standardized sequences to integrate many modern, advanced capabilities into a seamless calculation that the user controls from a single input file. Additional utility modules are provided primarily for post processing data generated from the analysis sequences for advanced studies. Input for SCALE sequences is provided in the form of text files using free-form input with extensive use of keywords and engineering-type input requirements. A GUI is provided to assist in the creation of input files, visualization of geometry and nuclear data, execution of calculations, viewing output, and visualization of results. An overview of the major SCALE capabilities and the analysis areas they serve is provided in Table 1, with additional descriptions provided below.

Summary of major SCALE capabilities




Analysis function(s)

Criticality Safety



3D multigroup and continuous energy eigenvalue Monte Carlo analysis and criticality search capability


Burnup credit analysis using 3D Monte Carlo


Hybrid 3D deterministic/Monte Carlo analysis with optimized fission source distribution

Reactor physics


1D and 2D general purpose lattice physics depletion calculations and generation of few-group cross section data for use in nodal core simulators

3D multigroup and continuous energy Monte Carlo depletion analysis

2D eigenvalue and reaction rate sensitivity analysis


2D streamlined light water reactor lattice physics depletion calculations and generation of few-group cross section data for use in nodal core simulators

Radiation Shielding


3D continuous energy and multigroup fixed-source Monte Carlo analysis with automated variance reduction

Activation, depletion and decay


General purpose point depletion and decay code to calculate isotopic concentrations, decay heat, radiation source terms, and curie levels


Simulated 2D and 3D analysis for light water reactor spent fuel assemblies (isotopic activation, depletion, and decay for light water reactor fuel assemblies)

ORIGEN reactor libraries

Pre-generated burnup libraries for a variety of fuel assemblies for commercial and research reactors

Sensitivity and uncertainty analysis


1D and 2D multigroup eigenvalue and reaction rate sensitivity analysis

3D multigroup and continuous energy eigenvalue and reaction rate sensitivity analysis

Determination of experiment applicability and biases for use in code and data validation


Stochastic uncertainty quantification in results based on uncertainties in nuclear data and input parameters

Material specification and Cross section processing


Temperature correction, resonance self-shielding, and flux weighting to provide problem-dependent microscopic and macroscopic multigroup cross section data integrated with computational sequences, but also available for stand-alone analysis

Standard composition library

Library used throughout SCALE that provides individual nuclides; elements with tabulated natural abundances; compounds, alloys, mixtures, and fissile solutions commonly encountered in engineering practice


3D Monte Carlo calculation of Dancoff factors

Monte Carlo transport



Eigenvalue Monte Carlo codes applied in many computational sequences for multigroup and continuous energy neutronics analysis


Fixed source Monte Carlo code applied in the MAVRIC sequence for multigroup and continuous energy analysis

Deterministic transport


1D discrete ordinates transport applied for neutron, gamma, and coupled neutron/gamma analysis


2D extended step characteristic transport with flexible geometry applied to neutronics analysis, especially within the TRITON sequences


3D Cartesian geometry discrete ordinates transport applied for neutron, gamma, and coupled neutron/gamma analysis, especially to generate biasing parameters within the MAVRIC and Sourcerer sequences (not generally run as stand-alone code in SCALE)

Nuclear Data

Cross Section Data

Recent neutron, gamma and coupled neutron/gamma nuclear data libraries in continuous-energy and several multigroup structures for use in all transport modules


Recent nuclear decay data, neutron reaction cross sections, energy-dependent neutron-induced fission product yields, delayed gamma ray emission data, neutron emission data, and photon yield data

Covariance Data

Recent uncertainties in nuclear data for neutron interaction, fission product yields, and decay data for use in TSUNAMI tools and Sampler



Numerous pre- and post-processing utilities for data introspection and format conversion