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Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHENIX End-of-Life Experiments
Closed for proposals
Project Type
Project Code
I33012CRP
1455Approved Date
Status
Start Date
Expected End Date
Completed Date
1 January 2013Description
The CRP addresses end-of-life experiments to be performed before the final shut-down of the prototype fast breeder power reactor PHENIX. The CRP will improve the participants’ analytical capabilities in the various fields of research and design of sodium cooled fast reactors through data and codes verification and validation.
Objectives
To improve the CRP participants analytical capabilities in various fields of research and design of sodium-cooled fast reactors through code verification and validation, with particular emphasis on temperature and power distribution calculations and the analysis of sodium natural circulation phenomena.
Specific objectives
Enhanced international team-building and technical cooperation in the area of fast reactors
Improved understanding of fast reactor neutronics and thermal hydraulics in safety relevant situations
Improved understanding of the methodology employed to simulate fast reactors (data and computer codes)
Improved verification and validation status of this methodology and the used numerical codes
Impact
The CRP contributed to improve the know-how on the power and reactivity control as well as some natural circulation phenomena occurring in a large sodium-cooled fast reactors, under transient and accident conditions. It was also very successful as far as the further verification, validation and qualification (V&V&Q) of the neutronic, safety and CFD numerical codes used by the CRP participants to assess performances and conduct safety analyses of innovative sodium-cooled fast reactors under development/design in their country.
Relevance
The outcomes of the CRP are of great importance for all the Member States with an active programme on fast reactors. As already stated, the CRP greatly contributed to improve the V&V&Q process of the most advanced nuclear codes used for the design and the safety analysis of innovative fast neutron systems under development/design worldwide.