Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

Moisseytsev, A. ; Sienicki, J. (Nuclear Engineering Division)

关键词: ACCIDENTS;    ANL;    BOILING;    BRAYTON CYCLE;    CONTROL ELEMENTS;    COOLANTS;    DESIGN;    ELECTRIC MOTORS;    FAST REACTORS;    FLOW RATE;    FREEZING;    HEAT EXCHANGERS;    LIQUID METALS;    MOTORS;    NUCLEAR POWER PLANTS;    REACTOR CORES;    SODIUM;    TRANSIENTS;   

DOI  :  10.2172/1011292
RP-ID  :  ANL-GENIV-147
PID  :  OSTI ID: 1011292
Others  :  TRN: US1102168

美国|英语

来源: SciTech Connect

【 摘 要 】

The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate inlet sodium temperatures by about 10 C. This temperature rise could presumably be precluded or significantly reduced through fine adjustment of the control rods and pump motors. The third option assumes that the reactor core power and primary and intermediate system flow rates are ideally reduced linearly in a programmed fashion that instantaneously matches the prescribed load demand. The calculated behavior of this idealized case reveals a number of difficulties because the control strategy for the S-CO{sub 2} cycle overcools the reactor potentially resulting in the calculation of sodium bulk freezing and the onset of sodium boiling. The results show that autonomous SFR operation may be viable for the particular assumed load change transient and deserves further investigation for other transients and postulated accidents.