The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models View Full Text


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Article Info

DATE

2018-05

AUTHORS

N. A. Mosunova

ABSTRACT

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium–plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal–hydraulic, neutronics, and thermal–mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal–hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code’s thermal–hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors. More... »

PAGES

304-316

References to SciGraph publications

  • 2013-05. Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors in THERMAL ENGINEERING
  • 2016-02. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems in THERMAL ENGINEERING
  • 2012-02. On a nuclear power strategy of Russia to 2050 in ATOMIC ENERGY
  • 2017-08. Neutron-Physical Model Impact on the Calculation of a Serious Accident with Sodium Boiling in a Fast Reactor in ATOMIC ENERGY
  • 2009-06. Modeling the diffusion yield of radioactive fission products from uranium dioxide fuel in ATOMIC ENERGY
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  • 1986-05. Analysis of the fast reactors' fuel-rod bundle flow resistance in SOVIET ATOMIC ENERGY
  • 2014-08. Russian Codes for Safety Analysis of Sodium-Cooled Fast Reactors in ATOMIC ENERGY
  • 2015-03. Heat-Exchange Models in the SOKRAT-BN Code for Calculating Sodium Boiling in Geometrically Different Channels in ATOMIC ENERGY
  • 2017-07. System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment in THERMAL ENGINEERING
  • 2013-06. Development of program modules with space-time kinetics for calculating unanticipated accidents in fast reactors in ATOMIC ENERGY
  • 2015-10. A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies in ATOMIC ENERGY
  • 2017-07. Modeling of Oxide Layer Formation and Corrosion Products Coagulation and Transport in Lead Coolant Using the OXID Module of the HYDRA-IBRAE/LM Code in ATOMIC ENERGY
  • Journal

    TITLE

    Thermal Engineering

    ISSUE

    5

    VOLUME

    65

    Author Affiliations

    Identifiers

    URI

    http://scigraph.springernature.com/pub.10.1134/s0040601518050063

    DOI

    http://dx.doi.org/10.1134/s0040601518050063

    DIMENSIONS

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