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Experts Meet at IAEA to Evaluate Computer Codes for Severe Accidents

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Employees at the main control room of the Smolensk Nuclear Power Plant in Russia. (Photo: Rosenergoatom)

This past week, 37 participants from 19 Member States attended a technical meeting on the Status and Evaluation of Severe Accident Simulation Codes for Water Cooled Reactors (WCRs).

Nuclear power plant safety systems are designed to mitigate a range of atypical operating conditions. Defined as “accident conditions more severe than a design basis accident” and “involving significant core degradation”, severe accidents are beyond design accidents – low probability but high impact. In these highly unlikely events, computer codes are used to model a wide range of associated phenomena – thermal-hydraulics, heating, hydrogen generation and combustion, reactor vessel failure, core melting, molten core–concrete interactions, containment performance, and fission products release.

Especially after the 2011 Fukushima Daiichi accident, nuclear experts have intensified the evaluation of severe accidents with increased attention to severe accident computational codes and modelling methods. By increasing the capacity of computation technology, safety experts have evaluated these codes to reduce uncertainties and improve calculation accuracy.

“This meeting was important since it allowed for different levels of interaction between end-users and developers,” said Ortiz Villafuerte, a nuclear engineer from Mexico. “We have been able to express our needs and will take meeting results back to our regulatory bodies.”

Yu Shen, nuclear engineer from the UAE also underlined: “We use these meetings, not only for the valuable exchange of ideas, but also to find ways in which we can use these computational tools to manage risk and increase reliability.”

Meeting presentations and discussions will be published as a Technical Document (TECDOC). “We hope that this comprehensive review of the status and progress in deterministic simulation1 codes, benchmark models, and how these codes can be improved with future research and development will help Member States to improve their preparedness for such unlikely events,” added Tatjana Jevremovic, the Scientific Secretary.

Common Severe Accident Code Examples

  • MELCOR:  MELCOR is a fully integrated, engineering-level computer code developed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission to model  the progression of severe accidents in nuclear power plants. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. MELCOR applications include estimation of severe accident source terms, and their sensitivities and uncertainties in a variety of applications.
  • MAAP:(Modular Accident Analysis Program): MAAP is an Electric Power Research Institute (EPRI) owned and licensed computer software that simulates the response of light water and heavy water moderated nuclear power plants for both current and Advanced Light Water Reactor (ALWR) designs.  It can simulate Loss-Of-Coolant Accident (LOCA) and non-LOCA transients for Probabilistic Risk Analysis (PRA) applications as well as severe accident sequences, including actions taken as part of the Severe Accident Management Guidelines (SAMGs).  There are several parallel versions of MAAP4 for BWRs, PWRs, CANDU designs, FUGEN design and the Russian VVER PWR design.
  • ASTEC: (Accident Source Term Evaluation Code): simulates all the phenomena that occur during a severe accident in a water-cooled nuclear reactor, from the initiating event to the possible release of radioactive products (the 'source term') outside the containment. ASTEC has been developed jointly over a number of years by the IRSN and its German counterpart, the Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS). 
  • MACCS: The MELCOR Accident Consequence Code System (MACCS) code is used to perform probabilistic offsite consequence assessments for hypothetical atmospheric releases of radionuclides. The code models atmospheric transport and dispersion, emergency response and long-term protective actions, exposure pathways, early and long-term health effects, land contamination, and economic costs. MACCS is used by U.S. nuclear power plant license renewal applicants to support the plant specific evaluation of severe accident mitigation alternatives as part of an applicant's environmental report for license renewal. MACCS is also used in severe accident mitigation design alternatives and severe accident consequence analyses for environmental impact statements for new reactor applications.
  • SCDAP/RELAP5: The Integral Severe Accident Analysis Code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system and reactor core during severe accidents as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. The coolant system behavior is calculated using a two-phase model allowing for unequal temperatures and velocities of the two phases of the fluid, and the flow of fluid through porous debris and around blockages caused by reactor core damage.
  • AC2:  AC2 is a code system developed by the Gesellschaft für Anlagen und Reaktorsicherheit (GRS gGmbH) and consists of three main modules: ATHLET, ATHLET-CD and COCOSYS. ATHLET is designed to simulate all the relevant phenomena during a design basis accident inside the reactor vessel. ATHLET-CD covers the nuclear accident stage with core degradation including fission product release and transport inside the cooling circuit and phenomena in the lower plenum. COCOSYS is developed for the simulation of all relevant phenomena, containment systems and conditions during the course of design basis accidents and severe accidents. 
  • SOCRAT: (System of Codes for Realistic Assessments of Severe Accidents) is a computer code intended for a coupled modeling of a wide range of thermohydraulic, physicochemical, thermomechanical and aerosol processes at all stages of accident progression, starting from initial event and up to corium release following the reactor vessel failure and consequent ex-vessel processes in containment. The code is essentially developed to model VVER NPPs. SOCRAT's field of application includes licensing support, design of safety systems, planning of experiments, PSA support, SAMG development and verification, crisis centers support, and education. During the Fukushima accident in 2011, SOCRAT was used as one of the numerical tools to support decision making about the need on whether or not to evacuate the Russian population of the Far-East.

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1A deterministic simulation is based on mathematical modelling in which outcomes are determined through known relationships among states and events, without room for random variation. In such models, a given input will always produce the same output, such as in a known chemical reaction.

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