Pure Lead Thermodynamic Properties in SIMMER-III Code: A Comparative Review and New Evaluation Proposal
Authors: Bellomo Alessandro , Massone Mattia , Gianfelici Simone , Morita Koji , Tarantino Mariano , Pesetti Alessio , Cossu Vittorio , Rineiski Andrei 
Published: November 2024

The lead-cooled fast reactors (LFRs) have been identified as one of the most promising technologies among the Generation IV candidates. Since 2000, ENEA has been supporting the core design, safety assessment, and technological development of innovative nuclear systems cooled by heavy liquid metals (HLM), in particular LFRs. Efforts are devoted to developing and validating computational tools for specific applications to HLM systems, ranging from neutronics codes, system and core thermal-hydraulic codes, computational fluid dynamics, and fuel pin performance codes, including their coupling. In this framework, the SIMMER-III code (S-III) has sparked particular interest, since it has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs) and is capable of simulating multi-phase, multi-component materials in multi-velocity field with coupled neutronics and thermo-hydraulics capabilities.

This paper presents a review of Lead (Pb) thermophysical properties (TPPs) models implemented into S-III. Detailed knowledge of HLM thermodynamic properties is needed for reactor design and modeling under representative normal and accidental conditions, and a review of the material properties is necessary before proceeding to a specific validation phase for its application to LFR systems. To ensure maximum flexibility, S-III TPP models are expressed with parametric functions, that is, polynomial equations for the liquid and solid phases. These models are designed to satisfy basic thermodynamic relationships among equation of state (EOS) variables over the entire temperature ranges. The EOS and TPP models for use in the accident analysis code are then completed by determining all the parameters related to each reactor-core material. The S-III validation process has been mainly focused on the technological development of the sodium-cooled fast reactor (SFR) program. The present work is inspired by this activity to carry out a similar verification and validation program focused on LFRs.

The results of the comparative analysis based on the most up-to-date and reliable sources for Pb available in the literature are shown. The comparison is focused on the properties of liquid lead. The TPPs considered are the density, dynamic viscosity, thermal conductivity, surface tension, and heat capacity. The comparative analysis reveals a significant discrepancy between the specific heat at constant pressure implemented in SIMMER and the literature. A new model for the heat capacity at constant pressure of liquid lead is proposed, based on the Gurvich (1991) relationship. Finally, a case study simulation is proposed to quantify the effect on the calculation results of the new heat capacity formulation.

This activity is preparatory to the validation phase that will take place during the experimental campaign that will be carried out at ENEA Brasimone Research Center in the CIRCE facility, concerning the steam generator tube rupture in LFRs.