Analysis and preliminary design of primary heat exchanger failure testing facility for lead-cooled fast reactors

This paper presents an assessment aimed at evaluating primary heat exchanger (PHE) failure of the Westinghouse Electric Company Lead-cooled Fast Reactor (LFR) and at designing a facility for testing phenomena involved in such failure. The system thermal-hydraulic code RELAP5/MOD3.3 was used to develop a transient analysis simulation at reactor scale. Because of RELAP5/MOD3.3's inability to mix working fluids, the steam injection effect was evaluated using the SIMMER-III code. The limits and strengths of both codes are highlighted throughout the paper. The reactor-scale steady-state results are in good agreement with the nominal operating condition. The transient results show that lead pool surface level variation and primary system pressurization during the PHE failure event are limited.The PHE failure testing facility was characterized, and a preliminary layout was developed. A separate-effects transient inside the vessel was analyzed with SIMMER-III and RELAP5/MOD3.3 runs. The simulation outcomes have provided useful data to inform subsequent design stages for the test facility. Different configurations of the facility have been assessed, highlighting the strengths and weaknesses of each design. The most important issue was identified to be lead pool swelling, reaching the vessel's lid and blocking the pressure relief vent. This poses a safety hazard that must be addressed and has been raised for resolution in subsequent design stages. The so-called V4 configuration is suggested as a starting point for further improvement of the facility. Furthermore, a smaller failure opening and lower lead level in the vessel are suggested.