**4.1 The simulation models developed for CEFR**

The core geometry has been modeled in its full 3D configuration keeping the heterogeneity of most of the present structures (e.g., the hollowed pellet geometry has been modeled). The SAs model starts from the region above the nozzle (the nozzle is not modeled) and reaches the corresponding head. Only few regions of the core, considered less relevant for the simulations, have been homogenized for the sake of simplification (e.g., SAs Handling head, Spring, etc.). The spacer wires have been homogenized with the corresponding cladding, to guarantee the conservation of the stainless-steel mass. **Figure 17(a)** shows the horizontal section of the core taken at a height of 105.1 cm from the bottom; the operation layout is shown, with all the fuel SAs already loaded in the core. **Figure 17(b)** shows a vertical section crossing the center of the core, along the x axis: the homogenized components are noticeable, mostly on the upper part of each assembly.

**Figure 17.** *Serpent Geometrical Core Model: (a) Horizontal Section; (b) Vertical Section.*

The Serpent 2.1.31 code [15] was used for all simulations. The ACE format cross-section libraries used have been processed at different temperatures, to match the specification of the various experiment. For the Work Packages (WPs) analysed in this work the data made available by SCK-CEN were used. All the used libraries were based on ENDF/B-VIII.0 [16]; a comparison with the library based on ENDF/B-VII.1 [17] has been performed for one test case of WP1 and has been reported in the following section.

For each experiment, both whole core geometry and material densities have been adjusted to the experimental temperatures, to consider expansion effects, and, therefore, the leakage variation. For most of the materials involved, the temperature adjusted parameters have been determined making use of the linear thermal expansion coefficients. For the sodium coolant density, the correlation provided in the technical specification has been used, while for the Helium gas, the density has been evaluated dividing the mass of gas at cold condition (at 20°C) by the fuel rods free volume available after the expansion (at hot zero power temperature) of the surrounding materials.

For all the experiments, which require a multiplication factor, the Serpent implicit k effective has been considered. In **Table 1** the cycle population, as well as the number of active and inactive cycles are reported.

For each experiment, both whole core geometry and material densities have been adjusted to the experimental temperatures, to consider expansion effects, and, therefore, the leakage variation. For most of the materials involved, the temperature adjusted parameters have been determined making use of the linear thermal expansion coefficients. For the sodium coolant density, the correlation provided in the technical specification has been used, while for the Helium gas, the density has been evaluated dividing the mass of gas at cold condition (at 20 °C) by the fuel


**259**

**Figure 18.**

**# of fuel SAs loaded**

**Table 2.**

**Rod position**

**RE2 [mm] keff Std.** 

72 70 0.99817 6.00E-

72 190 0.99854 6.00E-

*WP2, Comparison with Experimental Data, Rod Worth.*

(ENDF/B-VII.1)

*WP1, Comparison with Experimental Data.*

72 190

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

rods free volume available after the expansion (at hot zero power temperature) of the surrounding materials. For all the experiments, which require a multiplication factor, the Serpent implicit k effective has been considered. In **Table 1** the cycle population, as well as the number of active and inactive cycles are reported.

The results presented here are taken from [18]. First, the results of WP1 are analysed. **Table 2** shows the comparison between the calculated and measured values of the effective multiplication factor for the three supercritical steps and the critical position. The relative errors are small and practically constants for all the cases when ENDF/B-VIII.0 nuclear data are used, that suggests a systematic error originating from the nuclear data library can hold. Confirming this, a much better agreement is obtained when use is made of the ENDF/B-VII.1 nuclear data, as the error drops from 0.19% to 0.03% in the case of the RE2 positioned at 190 mm. Secondly the results of WP2 are presented. **Figure 18** shows the comparison between the calculated and measured integral worth of different control rod SAs, of the two-shutdown system (groups of control rod SAs, considered both fully

**Dev.**

05

72 151 0.99837 6.10E-05 1.000245 0.19% 72 170 0.99848 6.50E-05 1.000335 0.19%

05

**Serpent output Experimental Relative error**

1.00072 6.30E-05 0.03%

**keff,exp** k k / k 100 *eff eff eff* ,exp − ∗ ,exp

1.000000 0.18%

1.000395 0.19%

**4.2 Simulation results and preliminary interpretation of the CEFR SMs**

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

**Table 1.** *Simulations Set-up.* rods free volume available after the expansion (at hot zero power temperature) of the surrounding materials. For all the experiments, which require a multiplication factor, the Serpent implicit k effective has been considered. In **Table 1** the cycle population, as well as the number of active and inactive cycles are reported.
