Sodium-cooled Fast Reactor (SFR) Benchmark Task Force
1000 MW Metal Fuel SFR Image: OECD/NEA

Sodium-cooled fast reactors are the most promising type of reactors to achieve Generation IV (Gen IV) nuclear reactor goals at a reasonable time scale given the accumulated experience over the years. However, it is recognised that new regulations and safety rules as they exist worldwide are requiring improved safety performance. In particular, one of the foremost Generation IV International Forum (GIF) objectives is to design cores that can passively avoid core damage when the control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core.

The Sodium-cooled Fast Reactor (SFR) Benchmark Task Force was formed under the Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS). The mandate was proposed to work towards a shared analysis of the feedback and transient behaviour of next generation sodium-cooled fast reactor concepts.


A step-by-step analysis approach was proposed:

  1. Compile a state-of-the-art report which reviewed past and recent studies performed in the framework of sodium fast reactor and build a bibliographic repository which would stress core transient behaviours as a function of fuel characteristics (oxide, carbide, nitride and metal).
  2. Perform a parametric study based on two different core sizes: large size core (3600 MW thermal) and medium size core (1500-2500 MW thermal). For both cores sizes three types of fuel were proposed: oxide, carbide and metal. The comparative study was aimed at identifying the advantages and drawbacks for each concept based on nominal performances and global safety parameters:
    • Neutronics characterisation of global parameters (k-eff, power and flux distributions, void effect, Doppler, etc.)
    • Feedback coefficient evaluation, discussion and agreement on corresponding calculation methodology
    The group provided initial core descriptions for both size cores. All the input data were provided to the WPRS community and each contributor performed step-by-step core analysis and benchmark comparisons using best-estimate calculations.
  3. Based on the results obtained in the previous step, transient calculations were performed on a few selected cases for the principal unprotected transients (unprotected transient overpower (UTOP), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS) and the core behaviours characterised using a matrix classification.
  4. Synthesis of the whole work into a final report including recommendations to improve safety and future work toward severe accidents and minor actinides management.

The data were mainly geometry and materials definitions that enable neutronics characterisation of the cores for both equilibrium beginning and end of cycle states. No thermal-hydraulic data were provided.