The CRP addresses the natural convection behavior of the coolant in the reactor vessel of a sodium cooled fast reactor. The CRP participants will perform benchmark exercises focusing, in a first stage, on the numerical simulation of the sodium stratification measurements performed in the MONJU reactor vessel during the original start-up experiments. For the first stage of the CRP, the participants will analyse the sodium thermal stratification effects in the MONJU reactor vessel upper plenum after a plant trip test conducted in December 1995 with the reactor at 45% thermal power level simulating an abnormality in the condenser as triggering event
The overall objective of the CRP was to improve Member States’ analytical capabilities in the field of fast reactor in-vessel sodium thermal-hydraulics. A necessary condition towards achieving this objective is a wide international validation effort of the data and codes currently employed for the simulation of the various physical effects involved in this field. The CRP contributed towards achieving this objective with the help of benchmark exercises focusing on the numerical simulation of thermal stratification of sodium observed in the Monju reactor vessel at a turbine trip test conducted in December 1995 during the original start-up experiments, and with the help of a thorough assessment of the calculation versus measured data comparisons.
Identification of weaknesses in current methodologies and of the R&D needs to resolve the identified open issues
Validation of various multi-dimensional fluid dynamics codes in use in Member States through simulation of sodium cooled fast reactor upper plenum temperature distributions and comparison with measured data
The CRP allowed to improve the analytical capabilities of the participating organizations in the field of in-vessel sodium thermal-hydraulic, as well as to identify the key parameters which affect the moving-up rate of the thermal stratification in the upper part of the vessel of large sodium-cooled fast reactor. This know-how contributed to the validation and qualification of the simulation tools used by the organizations involved in the development and design of innovative SFRs worldwide. Another significant impact was the education and training of young nuclear specialists and PhDs who largely contributed to the different simulations and analyses.
This CRP is very relevant in the context of the overall effort of the project 1000154 aimed at increasing the know-how of the member states with an active fast reactor programme in the field of thermal-hydraulic and safety of sodium-cooled fast reactors. Moreover, this CRP is part of a set of CRPs aimed at V&V&Q of data and advanced methods and codes normally used for the design and the safety analysis of fast reactors. The other complementary CRPs are I33012 (PHENIX), I31021 (EBR-II, I31024 (NAPRO) and, in previous years, the CRP on BN-600. This large effort is possible thanks to the disclosure of costly experimental data by countries with FRs in operation. In particular the IAEA is thankful to JAERI which disclosed the results of the turbine trip test conducted in the MONJU reactor during the start-up campaign in 1995. In order to advance in this effort, in the near future it would be of paramount importance to gather other experimental data from MONJU, as well as from the operation of CEFR (China) and BN-800 (Russia). JAERI has already taken commitment to disclose new experimental data aimed at dynamic analysis of the whole primary heat transport system of the MONJU reactor. The relative benchmark exercise will be conducted in the frame of a new CRP planned within the P&B 2016-17 and further.