## 3 Simulation Codes and Methods and Salt Data  
The simulations for this feasibility study were performed using fast running unstructured mesh 2D deterministic simulations based on the SCALE/POLARIS neutron transport solver. POLARIS is a module of the SCALE code system [25] for 2D deterministic transport simulation, mainly developed for cross section preparation for core simulators. POLARIS is a new module of SCALE 6.2 that provides a 2D lattice physics analysis capability, using a multigroup self-shielding method called the Embedded Self Shielding Method (ESSM) and a transport solver based on the Method of Characteristics (MoC). In general, POLARIS and its cross section library have been developed and validated for light water reactors; thus, we have provided in a recent publication [26] the first verification against the Monte-Carlo code SERPENT [27] and some new, extensive follow up verifications, by comparison with different deterministic and statistic solvers of the SCALE package including some continuous energy Monte-Carlo comparisons [28]. The salt data for the applied fuel composition were based on two advanced Russian publications by [PERSON] and [PERSON] [29] and by [PERSON] and [PERSON] [30]. The base configuration is NaCl-UCl\({}_{3}\)-UCl\({}_{4}\) with the eutectic composition of 45% (molar) NaCl-38.5% (molar) UCl\({}_{4}\)-16.5% (molar) UCl\({}_{3}\). It is worthy of further detailed investigation due to its relatively high loading of fertile heavy metal material at the eutectic point, which makes it a very promising composition for achieving the required breeding efficiency essential for future self-sustained breeding [31]. Thus, it is an ideal candidate to be used for a first investigation of the fission product accumulation and its effect on criticality when reaching a significant burnup to identify elements for a potential future element-based clean-up system. However, in the longer term, this composition will require a large upgrading of the thermo-physical dataset to a much wider temperature range as input for any kind of future modelling and simulation studies of a potential reactor. At the mentioned eutectic point, the NaCl-UCl\({}_{3}\)-UCl\({}_{4}\) salt has a melting point of 338 \({}^{\circ}\)C [29] and a density of 3.232 g/cm\({}^{3}\) at 980 K.  
The model for this investigation, see Figure 2, is based on and developed from two recent publications [26; 31]. It has been developed for the evaluation of the burnup and breeding performance in a molten salt reactor as well as for the study of potential reactor control through temperature feedback. To allow for this, this very specific 2-D model has been built to take care of the very important effect of neutron leakage from a molten salt reactor core. The ring type model (thin black rings indicate the grid) consists of the reactor core of variable size (red), surrounded by a stainless-steel vessel of 20 mm thickness (blue) and a sodium reflector of 50 cm thickness (yellow). The arrangement is surrounded by a thin vacuum gap (light grey rings) and the rest of the unit cell is filled with a strong absorber (dark grey) to avoid the reflection of leaking neutrons from the core/reflector system through the reflective boundary conditions of the code originally developed to represent a unit cell in an infinite repetition of a fuel assembly. The simplified modeling approach here is to; calculate the fuel composition data accumulated to a chosen burnup (here 100 GWd/tHM) using a conventional reactor approach without feeding and cleaning so as to have a sufficient amount of fission products accumulated to be able to study concentrations and criticality effect of each element to a reasonable accuracy. This is a simplified model as the final system will require an online clean-up system as well as an online feeding system, and therefore, will not involve batch reprocessing at fixed burnup, but the nature and degree of separation will be correlated in these cases.