A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs; and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

This control switching device includes: a first operating panel provided at a first place and having A1, A2 buttons connected to a first control device and B1, B2 buttons connected to a second control device; and a second operating panel provided at a second place and having a1, a2 buttons connected to the first control device and b1, b2 buttons connected to the second control device. The first control device includes a first determination circuitry which determines whether to shut down input/output to/from an input/output device located at the first place, using operation signals of the A1, A2, B1, B2 buttons from the first operating panel. The second control device also includes a second determination circuitry similar to the first determination circuitry.

This control switching device includes: a first operating panel provided at a first place and having A1, A2 buttons connected to a first control device and B1, B2 buttons connected to a second control device; and a second operating panel provided at a second place and having a1, a2 buttons connected to the first control device and b1, b2 buttons connected to the second control device. The first control device includes a first determination circuitry which determines whether to shut down input/output to/from an input/output device located at the first place, using operation signals of the A1, A2, B1, B2 buttons from the first operating panel. The second control device also includes a second determination circuitry similar to the first determination circuitry.

A steam generator accident mitigation system is disclosed. A steam generator accident mitigation system to mitigate an accident if the accident occurs in a steam generator installed inside a containment building of a nuclear power plant according to an exemplary embodiment of the present system, the system including: a pressurizing tank which is installed inside the containment building and includes a first cooling water and a non-condensable gas for pressurizing the first cooling water therein; at least one connecting pipe connecting the steam generator and the pressurizing tank; and at least one connecting pipe valve which is installed in the at least one connecting pipe, respectively, and is able to control the amount of opening of the connecting pipe; wherein opening of the at least one connecting pipe valve permits fluid communication between the steam generator and the pressurizing tank.

A steam generator accident mitigation system is disclosed. A steam generator accident mitigation system to mitigate an accident if the accident occurs in a steam generator installed inside a containment building of a nuclear power plant according to an exemplary embodiment of the present system, the system including: a pressurizing tank which is installed inside the containment building and includes a first cooling water and a non-condensable gas for pressurizing the first cooling water therein; at least one connecting pipe connecting the steam generator and the pressurizing tank; and at least one connecting pipe valve which is installed in the at least one connecting pipe, respectively, and is able to control the amount of opening of the connecting pipe; wherein opening of the at least one connecting pipe valve permits fluid communication between the steam generator and the pressurizing tank.

An underground nuclear power reactor having a hollow blast tunnel which extends from one end of a containment member which houses a nuclear reactor, heat exchanger, generator, etc. A hollow blast tunnel extends from one end of the containment member with a normally closed door positioned therebetween. The blast tunnel defines a blast chamber having a plurality of spaced-apart debris deflectors positioned therein. The blast chamber has an upper wall with a roof opening formed therein which is selectively closed by a roof portion. If the reactor needs to be repaired or replaced, the door is opened so that the reactor will pass therethrough into the blast chamber and outwardly through the roof opening. If the reactor explodes, the blast therefrom drives the debris therefrom through the door and into the blast chamber where the deflectors reduce the blast force as the debris passes through the blast chamber.

An external reactor vessel cooling and electric power generation system according to the present invention includes an external reactor vessel cooling section formed to enclose at least part of a reactor vessel with small-scale facilities so as to cool heat discharged from the reactor vessel, a power production section including a small turbine and a small generator to generate electric energy using a fluid that receives heat from the external reactor vessel cooling section, a condensation heat exchange section 140 to perform a heat exchange of the fluid discharged after operating the small turbine, and condense the fluid to generate condensed water, and a condensed water storage section to collect therein the condensed water generated in the condensation heat exchange section, wherein the fluid is phase-changed into gas by the heat received from the reactor vessel. The external reactor vessel cooling and electric power generation system according to the present invention can continuously operate even during an accident as well as during a normal operation to cool the reactor vessel and produce emergency power, thereby enhancing system reliability. The external reactor vessel cooling and electric power generation system according to the present invention can easily apply safety class or seismic design using small-scale facilities, and its reliability can be improved owing to applying the safety class or seismic design.

An external reactor vessel cooling and electric power generation system according to the present invention includes an external reactor vessel cooling section formed to enclose at least part of a reactor vessel with small-scale facilities so as to cool heat discharged from the reactor vessel, a power production section including a small turbine and a small generator to generate electric energy using a fluid that receives heat from the external reactor vessel cooling section, a condensation heat exchange section 140 to perform a heat exchange of the fluid discharged after operating the small turbine, and condense the fluid to generate condensed water, and a condensed water storage section to collect therein the condensed water generated in the condensation heat exchange section, wherein the fluid is phase-changed into gas by the heat received from the reactor vessel. The external reactor vessel cooling and electric power generation system according to the present invention can continuously operate even during an accident as well as during a normal operation to cool the reactor vessel and produce emergency power, thereby enhancing system reliability. The external reactor vessel cooling and electric power generation system according to the present invention can easily apply safety class or seismic design using small-scale facilities, and its reliability can be improved owing to applying the safety class or seismic design.

A method to determine a global core reactivity bias and the corresponding estimated critical conditions of a nuclear reactor core prior to achieving reactor criticality. The method first requires collection and evaluation of the inverse count rate ratio (ICRR) data; specifically, fitting measured ICRR vs. predicted ICRR data. The global core reactivity bias is then determined as the amount of uniform reactivity adjustment to the prediction that produces an ideal comparison between the measurement and the prediction.

A method to determine a global core reactivity bias and the corresponding estimated critical conditions of a nuclear reactor core prior to achieving reactor criticality. The method first requires collection and evaluation of the inverse count rate ratio (ICRR) data; specifically, fitting measured ICRR vs. predicted ICRR data. The global core reactivity bias is then determined as the amount of uniform reactivity adjustment to the prediction that produces an ideal comparison between the measurement and the prediction.