Thermal-Hydraulic Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment With Modified RELAP5

Abstract

Sodium-cooled fast reactor (SFR) is one of most promising Generation IV reactor technology and has a rapid development in recent years. Experimental Breeder Reactor II (EBR-II) designed by Argonne National Laboratory (ANL) is a typical sodium-cooled fast reactor with a sodium-bonded metallic fuel core, featured with reactor negative reactivity feedback. In order to verify and validate (V&V) the inherent safety performance of sodium-cooled fast reactor, the International Atomic Energy Agency (IAEA) established the coordinated research project (CRP) in which Xi’an Jiaotong University has participated. The “Benchmark Analysis of EBR-II shutdown heat removal tests” was conducted by ANL. SHRT-45R, namely Unprotected Loss of Flow (ULOF), is one of the reactor tests among many transients. Argonne National Laboratory has prepared a detailed benchmark specification and has provided the enough benchmark data for SHRT-45R. At the beginning of SHRT-45R both primary main pumps and the intermediate loop pump were synchronously tripped to simulate the unprotected loss of flow accident. During the test, the plant protection system (PPS) was disabled to initiate a control rod scram. The SHRT-45R test demonstrated that EBR-II could keep in safe during the potentially adverse consequences of unprotected accidents. This paper introduces the models for predicting SHRT-45R in detail and presents the results of the analysis of the Unprotected Loss of Flow (ULOF) test SHRT-45R performed in the EBR-II reactor. The thermal-hydraulic calculations are performed with Modified RELAP5 in which the thermodynamic and transport properties of liquid and vapor state sodium have been supplemented, as well as the specific heat transfer correlations. The numerical results show that the loss of forced coolant flow causes the coolant temperatures in the instrumented subassemblies XX09 and XX10 to increase to a peak point but keep at an acceptable level about 930K and 850K at the early state of accidents, and then the reactor can shut down by itself due to the negative reactivity feedback including Doppler reactivity feedback and density reactivity feedback after 70s. The variation of the key thermal-hydraulic parameters including the coolant temperatures and fuel cladding temperature in the instrumented subassembly has a good agreement with the experimental data in general. The results could not only verify and validate the inhere safety performance of sodium-cooled fast reactors during potentially unprotected accidents but also demonstrate the analysis capability of the modified RELAP5 for sodium-cooled fast reactors.