**6. Conclusions**

*Recent Advances in Numerical Simulations*

these models are benchmarked against.

• SFRs use fuel pins tightly packed in a hexagonal duct. As said, to maintain a gap for coolant to flow through, fuel pins are separated with a metallic helical wire spacer wrapped around the pin entire length. Additionally, these wraps mitigate vortex-induced vibration and increase convective heat transfer by enhancing sub-channel mixing. Modeling the flow in wire-wrapped rod bundles is still a challenging problem [22]. Large uncertainties exist in the treatment of wire spacers and drag models used for momentum transfer in current low- resolution (lumped parameter) models. Sub-channel codes apply "forcing functions" to model wrap induced flow mixing. However, these approaches are limited to conditions submitted to a specific validation (flow regime, channel geometry, or operating conditions) and rely on complex coefficients which were derived from fitting the experimental databases the models are based on. High fidelity tools such as CFD can simulate wire-wrapped rod bundles with more detailed resolution. However, CFD simulations are still limited in their capability to characterize long-term transients or large system simulations. They are also bounded by the quality and resolution of the experimental data

• The axial power profile and the gamma heating outside the core region,

issue, thus improving the quality of computation results.

below and above the active fuel zone, affect the coolant temperature distribution. That way, they can have a significant impact on the behavior of some transients. In addition, approximations adopted in the calculation of the radial power distribution can engender large computation vs. measurement discrepancies for both dummy and reflector subassemblies (non-fuelled SAs). Accurately account for the gamma heating contribution to the power both in the active and non-active reactor regions can help to correctly address the

• The axial conduction heat transfer modelling is generally either non available or quite poor in the current SYS-TH codes. Even if the contribution of the axial heat conduction with respect to the other heat transfer mechanisms is always negligible in fast systems, accounting for the phenomenon in the heat structures improves the simulation results for the coolant temperature distribution outside the core region, while it does not provide any measurable gain for the evaluation of the cladding temperature, and even, sometimes it turns-out damaging. No specific action is recommended on

• The correct and comprehensive simulation of the heat transfer between adjacent subassemblies is mandatory to improve the agreement between computation results and measured data. Accounting for radial heat transfer with neighboring subassemblies is mandatory to obtaining good coolant temperature predictions. Accordingly, it is recommended to accurately account for the heat transfer between adjacent subassemblies and developing the suitable

• From a pure phenomenological point of view, the transport codes can catch all the phenomena included in the libraries. Nevertheless, data in the libraries have not got the same accuracy level for all phenomena, that can propagate discrepancies to the results. Accordingly, it is strongly recommended to carryout sensitivity and uncertainty analysis on the nuclear data, when estimated

**262**

this issue.

necessary.

computation capacity to do.

Sodium Fast Reactors is a branch of Fast Reactors technology developed since the early 60s, which nowadays regains large interest and attractiveness thanks to their flexibility and their potential to be operated as Actinide burners and/or breeders, thus playing a crucial role in the closure of nuclear fuel cycle and solving the burden of long-lived nuclear wastes.

Design and operation of such reactors require a noticeable computational capacity, but also specific means to assess their safety both in normal and downgraded operation as well as in emergency conditions. In this prospect, IAEA has organized several Coordinated Research Projects aimed at improving Member States' fast reactor analytical simulation capabilities.

The participation in such research programs - allowing direct comparison of computation results with measured data - contributes to increase the confidence in the capabilities of available computational tool, in the meantime highlighting the potential for improvements which could address and solve the pending issues and for the identification of new ones.

NINE has been actively involved in such research activities within the IAEA CRPs, thus catching the opportunity to independently assess and validate several commonly and widely used Thermal–Hydraulic and Reactor Physic codes. Moreover, it contributed to the comparison of results and the interpretation of discrepancy origins, identifying trends, and driving conclusions for future developments.

The NINE's Simulation Models have been developed strictly complying with the Best Estimate principle - which namely requires avoiding the introduction of inaccuracies due to rough approximations and assumptions - thus trying to represent at best the problem under investigation without adopting any major simplification. Despite this approach requires relevant and continuous computational efforts, the obtained results show-up highly satisfactory and encouraging in a whole. Moreover, as far as the computation capacity is concerned, the NEMM model methodology developed by NINE confirmed its applicability also in the case of SFR simulations.

The present chapter summarizes the activity carried-out, presents the results and discusses the main outcomes of the mentioned benchmark exercises, underlying the wide convergence among the computational tools, as well as detecting the main discrepancies and seeking for their common origin and trends, which should enable defining a mid-term vision for further development of the computer codes in the field of fast reactors and identifying new needs for their extended validation against either available or expected experimental data.

Among others, the following items have been identified as meriting careful and particular attention in the future: the need for an accurate modelling of the mixing of coolant flows in the assembly, the estimation of the pressure drop and the flow distribution during the transients which could be significantly improved using suitable empiric correlations, the sodium mixing and the thermal stratification phenomena which play a crucial role during the transients, and are sensitive to the nodalization scheme adopted and cannot be accurately predicted by the current existing SYS-TH codes, the need for a correct and comprehensive simulation of the heat transfer between adjacent subassemblies, the suitability for improvement of the calculation of inlet-plenum flow distribution.

Moreover, the importance of the representativeness, exhaustiveness and comprehensiveness of data have been once more and even farther pointed-out, claiming the need for complementing and/or implementing the existing data base with the results of experimental programs, engineered to match some specific validation needs, thus addressing, and filling the main and more crucial knowledge gaps. Definition of such programs should rely upon accurate to avoid duplication and dispersions.
