Simulation Methods

Energy Research, Inc. has been one of the leading organizations involved in the development of mathematical models and associated computer codes for
simulation of nuclear power plant systems and accidents.

Examples of advanced simulation models include:

  • Accident Diagnostics Analysis & Management System (ADAM-System ©) - Designed, developed and copyrighted by Energy Research, Inc., the ADAM-System is the first and most comprehensive multi-purpose, multi-functional simulation platform that is used to provide national nuclear regulatory organizations and the
    utility emergency response staff an easy to use tool to diagnose and prognoses accidents using on-line measured plant data, supplemental with much
    faster (i.e., several orders of magnitude) than real-time computer simulations. Currently, ADAM-systems are operational for the following nuclear power plants:
    • Beznau (Switzerland)
    • Leibstadt (Switzerland)
    • Gösgen (Switzerland)
    • Mühleberg (Switzerland)
    • Novovoronezh Unit 5 (Russian Federation)
    • Bohunice V2 (Slovakia)
    • Mochavce-1 (Slovakia)
    • Paks (Hungary)
    • Armenian Nuclear Power Plant (Armenia)
    • Sizewell-B (preliminary version) (UK)

    The ADAM-System uses a specifically-designed and developed Graphical User Interface (GUI) based on state-of-the-art software technology.

  • E R I In-Vessel Recovery (ERI IVR) - Designed and developed by Energy Research, Inc., as part of work performed for the U. S. Nuclear Regulatory
    Commission to provide a mechanistic capability for analysis of lower head integrity by external cooling, and to assess the likelihood of vessel failure.
    The code uses an advanced GUI that was developed by Energy Research, Inc. This code was used by Energy Research, Inc. as part of a study for the U.S.
    Nuclear Regulatory Commission (NRC), to assess the likelihood of the lower head failure in the Westinghouse AP1000 nuclear power plant
    (See NUREG/CR-6849 for more details)
  • ERPRA-ST - Designed and developed by Energy Research, Inc. to use an extremely efficient rate-dependent algorithm for calculation of radiological
    releases following severe accidents based on parametric aerosol removal and deposition models. The code uses an advanced Graphical User Interface (GIU)
    that was developed by Energy Research, Inc. ERPRA-ST is also an integral element of the Energy Research, Inc. Level-2 probabilistic safety assessment
    (PSA) methodology including a capability for uncertainty analysis.
  • ERPRA-BURN - Designed and developed by Energy Research, Inc. to enable stochastic analysis of hydrogen and carbon monoxide combustion and
    determination of the conditional containment failure probability, using the stress-strength interference concept and a stratified Monte Carlo sampling
    technique. ERPRA-BURN also uses an advanced GUI that was developed by Energy Research, Inc. It has been used in Level-2 PSA studies for more than 20
    nuclear power plants worldwide.
  • ERPRA-ROCKET - Designed and developed by Energy Research, Inc. to calculate the lift-off potential of a reactor pressure vessel subject to break
    formed at the lower head of user-specified size. The uncertainty analysis methodology implemented in ERPRA-ROCKET uses a simple Monte Carlo technique
    together with a full integration of the NRC-sponsored MELCOR code for analysis of reactor blow-down, reactor cavity pressurization, and vessel lift-off,
    including the impact of restraints, atmospheric drag, and other contributing parameters. ERPRA-ROCKET uses an advanced GIU that was also designed and developed
    by Energy Research, Inc.
  • ERI - PSAC - The Energy Research, Inc. Pool Swell Analysis Code (PSAC) is based on a mechanistic modeling of bubble growth and hydrodynamic behavior to predict the hydrodynamic response of a pressure suppression pool to a large break loss of coolant accident in boiling water reactors. The code includes a multi-vent clearing model, bubble dynamics (based on the solution to the Rayleigh equation), a semi-empirical bubble rise model, and the influence of pool inertial and other controlling parameters. ERI-PSAC is designed with a modern graphical user interface, and it has been successfully benchmarked against the General Electric Pool Swell Test Facility (PSTF) experimental data. The code was used as the basis of the approval of the Toshiba/Westinghouse GOTHIC analyses in support of the Combined Licensed Application for the South Texas units 3&4 nuclear power plants. The code is available upon request (approval for release by USNRC is required).