A method of operating a nuclear power system includes generating heat in a nuclear reactor core, transmitting the heat to a heat engine, generating electricity with a generator operatively coupled to the heat engine. The method further includes detecting a no-load condition, and stopping the heat engine. The method also includes transferring heat from an outer surface of the nuclear reactor to the environment through a heat transfer system if a temperature of the nuclear reactor rises above a threshold temperature. The method further includes preventing heat from transferring from the outer surface of the nuclear reactor to the environment through the heat transfer system if the temperature of the nuclear reactor is below the threshold temperature. Nuclear power systems and nuclear reactors are also disclosed.

A method of operating a nuclear power system includes generating heat in a nuclear reactor core, transmitting the heat to a heat engine, generating electricity with a generator operatively coupled to the heat engine. The method further includes detecting a no-load condition, and stopping the heat engine. The method also includes transferring heat from an outer surface of the nuclear reactor to the environment through a heat transfer system if a temperature of the nuclear reactor rises above a threshold temperature. The method further includes preventing heat from transferring from the outer surface of the nuclear reactor to the environment through the heat transfer system if the temperature of the nuclear reactor is below the threshold temperature. Nuclear power systems and nuclear reactors are also disclosed.

A method is for monitoring a nuclear reactor comprising a core in which fuel assemblies are loaded, each assembly comprising nuclear fuel rods each including nuclear fuel pellets and a cladding surrounding the pellets. The method includes determining (100) at least one operating time limit (T.sup.FPPI) for the extended reduced power operation of the reduced power nuclear reactor, so as to avoid a rupture of at least one of the claddings, operating (102) the nuclear reactor at reduced power for an actual time strictly less than the time limit (T.sup.FPPI), and relaxing (104) at least one threshold for protecting the nuclear power plant as a function of a difference between the time limit (T.sup.FPPI) and the actual time.

A method is for monitoring a nuclear reactor comprising a core in which fuel assemblies are loaded, each assembly comprising nuclear fuel rods each including nuclear fuel pellets and a cladding surrounding the pellets. The method includes determining (100) at least one operating time limit (T.sup.FPPI) for the extended reduced power operation of the reduced power nuclear reactor, so as to avoid a rupture of at least one of the claddings, operating (102) the nuclear reactor at reduced power for an actual time strictly less than the time limit (T.sup.FPPI), and relaxing (104) at least one threshold for protecting the nuclear power plant as a function of a difference between the time limit (T.sup.FPPI) and the actual time.

Provided in the present disclosure is a control system for a heat supply apparatus of a nuclear power plant, comprising: a first-stage pressure measurement means configured for measuring a first-stage pressure of a turbine to obtain a first-stage pressure signal; a high-exhaust pressure measurement means configured for measuring an exhaust pressure of a turbine high-pressure cylinder to obtain an exhaust pressure signal; a steam extraction heating flow rate measurement means configured for measuring a steam extraction heating flow rate to obtain a steam extraction heating flow rate signal; a data acquisition module configured for acquiring and transmitting the measured first-stage pressure signal, the measured exhaust pressure signal and the measured steam extraction heating flow rate signal to a core operation processing module; the core operation processing module; and a the signal output module.

Provided in the present disclosure is a control system for a heat supply apparatus of a nuclear power plant, comprising: a first-stage pressure measurement means configured for measuring a first-stage pressure of a turbine to obtain a first-stage pressure signal; a high-exhaust pressure measurement means configured for measuring an exhaust pressure of a turbine high-pressure cylinder to obtain an exhaust pressure signal; a steam extraction heating flow rate measurement means configured for measuring a steam extraction heating flow rate to obtain a steam extraction heating flow rate signal; a data acquisition module configured for acquiring and transmitting the measured first-stage pressure signal, the measured exhaust pressure signal and the measured steam extraction heating flow rate signal to a core operation processing module; the core operation processing module; and a the signal output module.

A nuclear reactor power system includes: a reactor core comprising a plurality of nuclear fuel elements, each nuclear fuel element comprising: a first cooling channel passing through the nuclear fuel element; and a second cooling channel passing through the nuclear fuel element and fluidly isolated from the first cooling channel; a first cooling system configured to transport a first fluid coolant through the reactor core, the first cooling system fluidly connected to the first cooling channel of each nuclear fuel element; and a second cooling system configured to transport a second fluid coolant through the reactor core, the second cooling system fluidly connected to the second cooling channel of each nuclear fuel element. A direction of first fluid coolant flow through the first cooling channel is the same as a direction of second fluid coolant flow through the second cooling channel.

A nuclear reactor power system includes: a reactor core comprising a plurality of nuclear fuel elements, each nuclear fuel element comprising: a first cooling channel passing through the nuclear fuel element; and a second cooling channel passing through the nuclear fuel element and fluidly isolated from the first cooling channel; a first cooling system configured to transport a first fluid coolant through the reactor core, the first cooling system fluidly connected to the first cooling channel of each nuclear fuel element; and a second cooling system configured to transport a second fluid coolant through the reactor core, the second cooling system fluidly connected to the second cooling channel of each nuclear fuel element. A direction of first fluid coolant flow through the first cooling channel is the same as a direction of second fluid coolant flow through the second cooling channel.

A nuclear reactor power system includes: a reactor core comprising a plurality of nuclear fuel elements, each nuclear fuel element comprising: a first cooling channel passing through the nuclear fuel element; and a second cooling channel passing through the nuclear fuel element and fluidly isolated from the first cooling channel; a first cooling system configured to transport a first fluid coolant through the reactor core, the first cooling system fluidly connected to the first cooling channel of each nuclear fuel element; and a second cooling system configured to transport a second fluid coolant through the reactor core, the second cooling system fluidly connected to the second cooling channel of each nuclear fuel element. A direction of first fluid coolant flow through the first cooling channel is the same as a direction of second fluid coolant flow through the second cooling channel.

A nuclear reactor power system includes: a reactor core comprising a plurality of nuclear fuel elements, each nuclear fuel element comprising: a first cooling channel passing through the nuclear fuel element; and a second cooling channel passing through the nuclear fuel element and fluidly isolated from the first cooling channel; a first cooling system configured to transport a first fluid coolant through the reactor core, the first cooling system fluidly connected to the first cooling channel of each nuclear fuel element; and a second cooling system configured to transport a second fluid coolant through the reactor core, the second cooling system fluidly connected to the second cooling channel of each nuclear fuel element. A direction of first fluid coolant flow through the first cooling channel is the same as a direction of second fluid coolant flow through the second cooling channel.