Primary pump coast-down characteristics analysis in lead cooled fast reactor under loss of flow transient

Abstract

The primary pump coast-down characteristics research is very important to ensure reactor safety during power failure events in the lead cooled fast reactor (LFR) design. The primary pump coast-down Flow Halving Time (FHT) effects on a 10 MWth lead-bismuth cooled research reactor with critical and sub-critical operation modes were analyzed under loss of flow transient in this paper. Comparison of different coast-down FHT effects on critical and sub-critical reactors was conducted. Moreover, other relevant parameters those may affect the coast-down FHT requirements (e.g. clad temperature limitation and temperature trigger threshold) were analyzed as well. The results showed that: Increasing the primary pump coast-down FHT can decrease the clad peak temperature in lead cooled fast reactor under LOF transient. Both unprotected and protected loss of flow (ULOF and PLOF) transients in critical reactor and ULOF transient in sub-critical reactor were insensitive to coast-down FHT, while PLOF transient in sub-critical reactor was found to be sensitive to the coast-down FHT. The coast-down FHT effects on sub-critical reactor was bigger than critical reactor in the early phase of LOF transient. The primary pump coast-down FHT had little effects on the critical reactor safety under LOF transients. The clad peak temperature of the sub-critical reactor under ULOF transient was the only one that exceeded the temperature limitation among these four transients. Raising the clad temperature limit of the sub-critical reactor contributes little to reactor safety under ULOF transient. Measures should be taken to avoid the happening of serious accident, such as ULOF transient in sub-critical reactor. Scram signal setup (e.g. core outlet temperature trigger threshold) would not enhance the safety of lead cooled critical and sub-critical reactors and would not change the coast-down FHT requirements under LOF transient.