**5. Challenges and opportunities for enhanced computation capacity and extended investigation in the sodium-cooled fast reactors field**

The core of SFR is subdivided into several hexagonal fuel bundles. Unlike LWR squared subassemblies, these hexagonal subassemblies are compact in size leading to higher power density. Higher power densities and higher coolant temperature can lead to coolant boiling. Besides the difference between the boiling points of the two coolants, a relevant distinguishing feature of sodium is its higher thermal conductivity. While the conduction in water is usually neglected when modeling LWRs, the heat conduction in sodium cannot be ignored in SFRs, mainly when natural circulation occurs. The difference in opacity between the two coolants, also implies adoption of different surveillance methods and the differences in the activation products lead to specific features of shielding (these topics are only mentioned here without deepening because they fall out the scope of this chapter).

Moreover, while LWRs systematically use uranium-oxide fuel (UO2), SFRs can use both oxide and metallic fuel, depending on the design features. The oxide fuel has the advantage of having a high melting temperature relative to the metallic. It is also less ductile and have higher strength. The prediction of the temperature distribution in fuel assemblies should be accurate to assure the safety and reliability of the reactor operation. The focus of the new codes and modeling will be on enhancements to the thermal hydraulics modeling aspects of the SFR and the modeling of metallic fuel.

**261**

*International Benchmark Activity in the Field of Sodium Fast Reactors*

design and the safety assessment of the Sodium Fast Reactors.

SFR computation capacity and reliability are the following:

Another key difference between the designs of Pressurized Water Reactors (PWRs) and SFRs is the structural support of the fuel. While the PWR fuel is supported by grid spacers that are located at specific heights, the SFR fuel is supported by wire-wrappings that extend along the whole length of the fuel. The wire wrapping brings the advantage of a better coolant flow mixing and a lower temperature gradient across the subassemblies, but that comes at the expense of higher pressure-

Accounting for the above-mentioned general considerations and trends as well as of the outcomes of the investigations carried-out in the framework of the Benchmark exercises described here above, in paragraphs 2 to 4, the Participants were able to drive some general conclusions on the expected new features of the computation tools as well as on the up-dated methodologies to be adopted for the

These conclusions identify the main trends for improvement. They can either contribute to further and proficiently expanding the computation capacity of existing tools (and even considering the development of new ones), as well as adapting and/or updating the methodologies adopted so far in the studies.

Moreover, the importance of the representativeness, exhaustiveness and comprehensiveness of the data stored in the data base adopted for validation of the computation tools have been once more and even farther pointed-out. Accordingly, it is considered worth complementing and/or implementing the existing data base with the results of ad-hoc experimental programs, accurately designed, and engineered to match some specific validation needs, thus addressing, and filling the main and more crucial knowledge gaps. Definition of such experimental programs should be made relying upon accurate investigation adopting, e.g., Kriging-like methodologies

The main topics found-out to be potential levers for further improvement in the

• A major need for the SFR M&S (Modelling and Simulation) appears to be the improvement of the calculation of inlet-plenum flow distribution and of lateral mixing of coolant flows in the assembly to reduce conservatism in the estimation of the peak fuel and cladding temperatures [19]. The high-fidelity Computational Fluid Dynamics (CFD) models should be developed, tested, and adapted for application to liquid sodium for the relevant flow characteristics and geometric configurations, especially wire-wrapped pin bundles contained within the hexagonal assembly boxes. Additional near-term needs for the SFR M&S are related to the thermal-mechanical modeling capabilities. Recent advances in modeling of oxide fuels behavior in thermal reactors (e.g., fuel and clad conductivity and gap conductance modeling) should also be adapted for fast reactor applications.

• The pressure drop and flow distribution estimation during the transients can be significantly improved when using suitable empiric correlations to model friction losses in the wire-wrapped fuel bundle region, as proposed by Cheng

• The sodium mixing and the thermal stratification phenomena play a crucial role during the transients, mainly during the natural circulation phase. They cannot be accurately predicted by the current existing SYS-TH (System Thermal–Hydraulic) codes. In addition, these phenomena are also sensitive to the nodalization scheme adopted in the calculations. It is suggested to address this major issue for design and operation through either suitable sensitivity analyses or more in-depth investigations (i.e., adopting CFD codes).

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

losses along the of the fuel height.

to avoid duplication and dispersions.

& Todreas [20] and Pontier [21].

#### *International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*

*Recent Advances in Numerical Simulations*

operating or with one SA stuck) and of all control rods together. The experimental values are obtained through a rod drop experiment. It appears from the figure that a noticeable good agreement between results and measurements has been

*WP6, Comparison with Experimental Measurements, Axial Reaction Rates.*

The last results belong to WP6. In **Figure 19** is reported the comparison between the reaction rates axial distributions evaluated with Serpent and the measured values. The agreement between measurements and simulations is generally quite good. Only the case of 197Au(n,γ) shows a noticeable difference particularly for the positions at the top and bottom of the core. Further investigating and understanding the origin and nature of this discrepancy would contribute to a better understanding of

**5. Challenges and opportunities for enhanced computation capacity and extended investigation in the sodium-cooled fast reactors field**

The core of SFR is subdivided into several hexagonal fuel bundles. Unlike LWR squared subassemblies, these hexagonal subassemblies are compact in size leading to higher power density. Higher power densities and higher coolant temperature can lead to coolant boiling. Besides the difference between the boiling points of the two coolants, a relevant distinguishing feature of sodium is its higher thermal conductivity. While the conduction in water is usually neglected when modeling LWRs, the heat conduction in sodium cannot be ignored in SFRs, mainly when natural circulation occurs. The difference in opacity between the two coolants, also implies adoption of different surveillance methods and the differences in the activation products lead to specific features of shielding (these topics are only mentioned here without deepening because they fall out the scope of this

Moreover, while LWRs systematically use uranium-oxide fuel (UO2), SFRs can use both oxide and metallic fuel, depending on the design features. The oxide fuel has the advantage of having a high melting temperature relative to the metallic. It is also less ductile and have higher strength. The prediction of the temperature distribution in fuel assemblies should be accurate to assure the safety and reliability of the reactor operation. The focus of the new codes and modeling will be on enhancements to the thermal hydraulics modeling aspects of the SFR and the modeling of

**260**

metallic fuel.

chapter).

achieved.

**Figure 19.**

the of the measured activity.

Another key difference between the designs of Pressurized Water Reactors (PWRs) and SFRs is the structural support of the fuel. While the PWR fuel is supported by grid spacers that are located at specific heights, the SFR fuel is supported by wire-wrappings that extend along the whole length of the fuel. The wire wrapping brings the advantage of a better coolant flow mixing and a lower temperature gradient across the subassemblies, but that comes at the expense of higher pressurelosses along the of the fuel height.

Accounting for the above-mentioned general considerations and trends as well as of the outcomes of the investigations carried-out in the framework of the Benchmark exercises described here above, in paragraphs 2 to 4, the Participants were able to drive some general conclusions on the expected new features of the computation tools as well as on the up-dated methodologies to be adopted for the design and the safety assessment of the Sodium Fast Reactors.

These conclusions identify the main trends for improvement. They can either contribute to further and proficiently expanding the computation capacity of existing tools (and even considering the development of new ones), as well as adapting and/or updating the methodologies adopted so far in the studies.

Moreover, the importance of the representativeness, exhaustiveness and comprehensiveness of the data stored in the data base adopted for validation of the computation tools have been once more and even farther pointed-out. Accordingly, it is considered worth complementing and/or implementing the existing data base with the results of ad-hoc experimental programs, accurately designed, and engineered to match some specific validation needs, thus addressing, and filling the main and more crucial knowledge gaps. Definition of such experimental programs should be made relying upon accurate investigation adopting, e.g., Kriging-like methodologies to avoid duplication and dispersions.

The main topics found-out to be potential levers for further improvement in the SFR computation capacity and reliability are the following:


**263**

developments.

collaborations.

**6. Conclusions**

of long-lived nuclear wastes.

reactor analytical simulation capabilities.

for the identification of new ones.

*International Benchmark Activity in the Field of Sodium Fast Reactors*

• The cross section pre-processing routine implemented in some codes, such as the Doppler-broadening pre-processor routine inside Serpent 2, are not able to adjust the temperature of the unresolved region probability tables. According to the importance of that region for the stability and operability of the core, when using such codes for fast reactor systems, it is recommended to evaluate

• When using Monte-Carlo codes, the calculation time to achieve a good statistic can turn-out remarkably high. This issue being in common with all Monte Carlo calculations and not being a specific problem for the Sodium Fast

• Many experiments have been performed to study thermal- hydraulics characteristics, primarily pressure drops, of the wire-wrapped fuel bundles. Often however, the uncertainties of pressure drop measurements associated with these experiments is high due to the geometrical complexity of the hexagonal wire-wrapped fuel bundle. Recently [23, 24], a database of pressure drops and flow-field measurements in a 61-pin wire-wrapped hexagonal fuel bundle was developed with the sole purpose to benchmark the existing correlations and validate the CFD calculations. These experiments investigate the flow characteristics in the near-wall region of the 61- pin wire-wrapped hexagon fuel bundle.

All the above-mentioned topics merit for careful consideration and further investigation in view of the definition of future R&D programs in support to the industrial development and the deployment of the SFRs. Due to their size and scope, they should mainly be addressed in the framework of international

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

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

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

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

the Doppler broadening with a suitable nuclear data processing code.

Reactor simulation, no specific recommendation is done.

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

*International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*


All the above-mentioned topics merit for careful consideration and further investigation in view of the definition of future R&D programs in support to the industrial development and the deployment of the SFRs. Due to their size and scope, they should mainly be addressed in the framework of international collaborations.
