Control characteristics of the steady-state operation scheme for coolant temperature in the primary loop of the lead-bismuth cooled fast reactor

<bold>Background</bold>The lead-bismuth cooled fast reactor (LBFR) exhibits strong nonlinearity, time-varying dynamics, and tight coupling, demanding precise control for safe operation. Designing control strategies requires establishing steady-state operation schemes <italic>via</italic> primary/secondary loop energy balance analysis, followed by evaluating their distinct control response behaviors under varying operational conditions.<bold>Purpose</bold>This study aims to obtain the control response characteristics of the nuclear steam supply system (NSSS) under the three different steady-state operation schemes: constant average coolant temperature of the primary loop, programmed operation of the outlet coolant temperature, and constant coolant temperature of the core inlet.<bold>Methods</bold>Firstly, the SVBR 100 NSSS was taken as the research object, and its simulation model based on the mass-momentum and energy conservation principles was established, including: reactor core model, hot and cold lead-bismuth pool models, once through steam generator model, pump and valve model, <italic>etc</italic>. Then, based on the developed simulation model of NSSS for LBFR, simulation analysis of typical conditions was carried out, including: 1) ±10% full power (FP) load step change; 2) ±5% FP·min ^-1 load linear change; 3) load rejection condition. Finally, the control response characteristics of the system under the three schemes of constant primary coolant average temperature, core outlet coolant temperature program operation and constant core inlet coolant temperature was comparatively investigated.<bold>Results</bold>Simulation results demonstrate that the three schemes can meet the stability and safety requirements of operation control and have certain control performance. The transient results of ±10% FP load step change show that the relative nuclear power overshoot is the smallest under the inlet coolant temperature constant control scheme; the nuclear power overshoot is the largest under the programmed core outlet coolant temperature control scheme, and the overshoot is 3% of the relative nuclear power. Due to the slower load linear change of ±5% FP·min ^-1, the three control schemes have good control effect on nuclear power and steam pressure, and the maximum steam pressure overshoot is 0.3 MPa. The load rejection results show that the coolant average temperature constant control scheme has a moderate control effect, compared with the inlet temperature control scheme, with smaller overshoot and faster regulation time; compared with the outlet temperature control scheme, it is more superior in the control rods position, and the rods insertion time is shorter. During the above transients, the lowest temperature of lead-bismuth is over 335 ℃, while the melting point of lead-bismuth is 125 ℃, and the safety margin is about 210 ℃; while the highest temperature of lead-bismuth during the transient process is about 352 ℃, which is much lower than its boiling point of 1 653.85 ℃. Therefore, the temperature response of the primary loop coolant obtained from the three control schemes has a large safety margin, and it is not necessary to worry about the solidification of lead bismuth when designing the control scheme, but should pay more attention to the system stability and the overshoot of the control temperature to reduce the thermal shock to the primary loop system.<bold>Conclusions</bold>Results of this study demonstrate that the constant core inlet temperature control scheme has excellent performance in relative nuclear power control, the programmed outlet coolant temperature control scheme has better performance in the coolant temperature response, and the constant average coolant temperature scheme has intermediate control performance.