FRENETIC

The design of innovative generation IV reactors, such as the Lead-cooled Fast Reactor, requires stability and safety assessments to be performed in both operational and accidental transient conditions. These studies typically require the use of multi-physics models. The Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code has been developed since 2011 with the ambition to provide (approximate) solutions for liquid metal-cooled fast reactor core design and/or safety analysis in a computationally effective (i.e. relatively fast) way, thanks to the fact that the 3D problem is solved with a simplified approach.

Reference geometry is the closed hexagonal fuel element configuration, as currently proposed for ALFRED prototype reactor within the framework of the FALCON consortium. The tool implements coupled neutronic (NE) and thermal-hydraulic (TH) models.

Main outputs of the code are the TH variables and power 3D distributions in the core at selected times, together with their inlet and outlet values during the whole transient.

In the NE module, a 3D full-core multi-group diffusion solver has been developed, based on a 3D coarse-mesh nodal. A photon transport model and a decay-heat model are also included.

In steady state mode, iterations are performed to compute the effective multiplication factor of the system.

In the transient mode three options are available:

-        the point kinetics method

-        a full discretisation of the prompt neutron balance equation and delayed neutron precursors balance equations with the one-step theta method

-        the quasi static method.

The NE module can be run in standalone mode to compute neutron flux distribution for a given input temperature.

In the TH module, the hexagonal elements, described by 1D (axial) transient advection and conduction in the coolant coupled to conduction in the fuel pins, are thermally coupled (in explicit, thanks to the weak thermal coupling between assemblies) to each other in the transverse directions to obtain the full-core evolution of the distribution of the TH variables.

Space derivatives along the axial direction are approximated by 1D linear finite elements, equivalent to central difference approximation. The set of equations for each assembly is solved implicitly in time (fully implicit or Crank-Nicolson schemes).

Also the TH module can be run in standalone mode to compute TH variables distribution and evolution in the core (or in a single channel) for a given power generation.

The two modules are coupled by transferring to the TH module the distribution of the power source computed by the NE module, which is the driver of the TH evolution; alternately, the temperature distribution computed by the TH module is input at each time step to the NE module in order to update the cross sections.

Research topics

• Multiphysics of advanced reactors

Publications

1. Full-core coupled neutronic/thermal-hydraulic modelling of the EBR-II SHRT-45R transient
   Article
   

   Caron, Dominic; Bonifetto, Roberto; Dulla, Sandra; Mascolino, Valerio; Ravetto, Piero; Savoldi, Laura; Valerio, Domenico; Zanino, Roberto
   INTERNATIONAL JOURNAL OF ENERGY RESEARCH
   John Wiley & Sons
   Vol.42 pp.17 (pp.134-150) ISSN:0363-907X DOI:10.1002/er.3571

1. New aspects in the implementation of the quasi-static method for the solution of neutron diffusion problems in the framework of a nodal method
   Article
   

   Caron, Dominic; Dulla, Sandra; Ravetto, Piero
   ANNALS OF NUCLEAR ENERGY
   ELSEVIER
   Vol.87 pp.15 (pp.34-48) ISSN:0306-4549 DOI:10.1016/j.anucene.2015.02.035

1. A full-core coupled neutronic/thermal-hydraulic code for the modeling of lead-cooled nuclear fast reactors
   Article
   

   Bonifetto, Roberto; Dulla, Sandra; Ravetto, Piero; Savoldi, Laura; Zanino, Roberto
   NUCLEAR ENGINEERING AND DESIGN
   Elsevier
   Vol.261 pp.10 (pp.85-94) ISSN:0029-5493 DOI:10.1016/j.nucengdes.2013.03.030

1. Full-Core Coupled Neutronic/Thermal-Hydraulic Model of Innovative Lead-Cooled Fast Reactors
   Proceeding
   

   Bonifetto, Roberto; Dulla, Sandra; Ravetto, Piero; Savoldi, Laura; Zanino, Roberto
   TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY
   In: TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY
   American Nuclear Society
   2012 ANS Annual Meeting (Chicago, IL (USA)) June 24-28, 2012
   Vol.106 pp.3 (pp.630-632) ISSN:0003-018X

2. PROGRESS IN MULTI-PHYSICS MODELING OF INNOVATIVE LEAD-COOLED FAST REACTORS
   Article
   

   Bonifetto, Roberto; Dulla, Sandra; Ravetto, Piero; Savoldi, Laura; Zanino, Roberto
   FUSION SCIENCE AND TECHNOLOGY
   American Nuclear Society
   Vol.61 pp.5 (pp.293-297) ISSN:1536-1055 DOI:10.13182/FST12-A13435

Total: 5