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Title: SCALE Code System 6.2.1

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 Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and 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 and three Monte Carlo radiation transport solvers that are selected based on the 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 graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.
Authors:
 [1] ;  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
OSTI Identifier:
1326509
Report Number(s):
ORNL/TM--2016/352
453040374; ORNL/TM-2005/039, V 6.2.1
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Research Org:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; 97 MATHEMATICS AND COMPUTING; 98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL PROTECTION; S CODES; REACTOR SAFETY; SAFETY ANALYSIS; ORNL; NEUTRONS; GAMMA RADIATION; SPENT FUELS; MONTE CARLO METHOD; COMPUTERIZED SIMULATION; DESIGN; SENSITIVITY ANALYSIS; SHIELDING; MATHEMATICAL SOLUTIONS; DATA COVARIANCES; SOURCE TERMS; RADIATION TRANSPORT; CRITICALITY; DECAY; DISPLAY DEVICES; MULTIGROUP THEORY; DATA VISUALIZATION; REACTOR LATTICES; REACTOR PHYSICS