Analysis of protected RIA and LOFA in plate type research reactor using coupled neutronics thermal-hydraulics system code

Marat Margulis, Erez Gilad

Research output: Chapter in Book/Report/Conference proceedingConference contributionpeer-review

4 Scopus citations

Abstract

The application of Best Estimate (coupled neutron kinetics/thermal-hydraulics - NK/TH) codes for research reactors safety analyses has gained considerable momentum during the past decade. This activity is largely facilitated by the high level of technological maturity and expertise attained by these techniques as NPPs safety technology and is largely driven by IAEA activities. The present study belongs in this framework, where a coupled NK/TH code (THERMO-T) was developed and applied to the analysis of protected reactivity insertion (RIA) and loss of flow (LOFA) accidents in a typical research reactor with standard MTR plate type fuel assemblies. The coupling is realized by considering the neutronic reactivity feedbacks of the fuel and coolant temperatures and a heat generation model for the reactor power. The neutron flux in the reactor core is solved by applying the point reactor kinetic equations, using radial and axial power distributions calculated from a 3D full core model by the three-dimensional continuous-energy Monte Carlo reactor physics code Serpent. The evolution of temporal and spatial distributions of both fuel and coolant temperatures is calculated for all fuel channels using a finite volumes time implicit numerical scheme for solving a three conservation equations model. In this study, three different thermal hydraulic models of the code are evaluated, as well as its sensitivity to different heat transfer correlations.

Original languageEnglish
Title of host publicationInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
PublisherAmerican Nuclear Society
Pages8796-8808
Number of pages13
ISBN (Electronic)9781510811843
StatePublished - 1 Jan 2015
Event16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015 - Chicago, United States
Duration: 30 Aug 20154 Sep 2015

Publication series

NameInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
Volume10

Conference

Conference16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015
Country/TerritoryUnited States
CityChicago
Period30/08/154/09/15

Keywords

  • Coupled system code
  • LOFA
  • RIA
  • Research reactor
  • Safety analysis
  • Transients

ASJC Scopus subject areas

  • Instrumentation
  • Nuclear Energy and Engineering

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