Page 210 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

System thermal hydraulics for liquid metals 181

The CATHARE code has already been modified in the recent past to treat several
other fluids (Geffraye et al., 2009). Within a specific topic of cooperation between
ENEA and CEA, the lead and lead-bismuth eutectic (LBE) thermodynamic properties
(OECD/NEA, 2015) have been implemented (Polidori, 2010) and now made available
to CATHARE users.

4.5.2 ATHLET code/package
The Analysis of Thermal Hydraulics of LEaks and Transients (ATHLET) package is
part of a suite of codes covering the simulation of all operational states, incidents, acci-
dents, and severe accidents in a nuclear power plant (ATHLET User’s Manual, 2016).
ATHLET is being developed for the analysis of anticipated and abnormal plant tran-
sients, small and intermediate leaks, and large breaks in light-water reactors. The aim
of the code development is to cover the whole spectrum of design-basis and beyond-
design-basis accidents (without core degradation) for PWRs and BWRs with only one
program.
ATHLET is composed of several basic modules for the simulation of the different
phenomena involved in the operation of a light-water reactor. The system configura-
tion to be simulated is modeled by connecting basic thermofluid dynamic elements,
called thermofluid and heat conduction objects. Within ATHLET, the time advance-
ment is performed by a forward-Euler/backward-Euler solver, which is a general-
purpose solver for the solution of nonlinear ordinary differential equation systems
of first order. ATHLET is being used by numerous institutions in Germany and
abroad. The code is applicable not only for western reactor designs but also for Rus-
sian reactors. Furthermore, ATHLET was extended to simulate Generation IV reactor
concepts by the application of working fluids like lead, LBE, and sodium. The used
version for this application is ATHLET 3.1A, Patch2. The main code features are:
– advanced thermal hydraulics;
– modular code architecture;
– separation between physical models and numerical methods;
– proprietary pre- and postprocessing tools;
– portability.

4.5.3 SPECTRA code
Sophisticated Plant Evaluation Code for Thermal-hydraulic Response Assessment
(SPECTRA) is a thermal-hydraulic system code developed by NRG, designed for
thermal-hydraulic analysis of nuclear power plants (Stempniewicz, 2016). The code
is applicable to light-water reactors (LWRs), liquid-metal-cooled fast reactors
(LMFRs), high-temperature reactors (HTRs), and molten-salt-fueled reactors
(MSRs). The code can be used for thermal accident scenarios involving loss-of-
coolant accidents (LOCAs), operational transients, and other accident scenarios in
nuclear power plants. Models include multidimensional two-phase flow; non-
equilibrium thermodynamics; transient heat conduction in solid structures; and a