A method for monitoring changes in the boron concentration in the coolant of a reactor during a nuclear plant outage that applies temperature compensation to the source range detector output. The method then monitors the compensated output signal to identify changes in the detector count rate above a preselected value.

A method for monitoring changes in the boron concentration in the coolant of a reactor during a nuclear plant outage that applies temperature compensation to the source range detector output. The method then monitors the compensated output signal to identify changes in the detector count rate above a preselected value.

An experimental system a method for studying jet impact characteristics at a core outlet of a fast reactor are provided. The system includes a jet impact main loop including a water storage tank, plunger pumps, a filter, preheaters, a jet impact chamber, a heat regenerator, a condenser, valves, flow meters and pipelines connecting these facilities; a cooling loop including cooling tower, a cooling pump, a regulating valve and a flow meter; and a makeup water loop including a deionized water machine, a makeup water tank and a plunger pump. Water in the water storage tank flows to the heat regenerator via the plunger pump, is preliminarily heated by the heat regenerator and then is divided into three branches to flow to the jet impact container.

An experimental system a method for studying jet impact characteristics at a core outlet of a fast reactor are provided. The system includes a jet impact main loop including a water storage tank, plunger pumps, a filter, preheaters, a jet impact chamber, a heat regenerator, a condenser, valves, flow meters and pipelines connecting these facilities; a cooling loop including cooling tower, a cooling pump, a regulating valve and a flow meter; and a makeup water loop including a deionized water machine, a makeup water tank and a plunger pump. Water in the water storage tank flows to the heat regenerator via the plunger pump, is preliminarily heated by the heat regenerator and then is divided into three branches to flow to the jet impact container.

A method, apparatus, and system according to which first and second transducers are connected to first and second waveguides, respectively, the first and second waveguides are connected to a pipe, and ultrasonic wave signals are exchanged between the first and second transducers, said ultrasonic wave signals passing through the first and second waveguides, the pipe, and a fluid in the pipe. A temperature of the fluid flowing in the pipe may exceed about 600° C. The first and second waveguides insulate the first and second transducers from the pipe and propagate the ultrasonic wave signals between the pipe and the first and second transducers, respectively, so that the ability of the first and second transducers to exchange the ultrasonic wave signals is not adversely affected by the temperature of the fluid in the pipe. The first and second waveguides may be made of a calcium silicate technical ceramic.

A method, apparatus, and system according to which first and second transducers are connected to first and second waveguides, respectively, the first and second waveguides are connected to a pipe, and ultrasonic wave signals are exchanged between the first and second transducers, said ultrasonic wave signals passing through the first and second waveguides, the pipe, and a fluid in the pipe. A temperature of the fluid flowing in the pipe may exceed about 600° C. The first and second waveguides insulate the first and second transducers from the pipe and propagate the ultrasonic wave signals between the pipe and the first and second transducers, respectively, so that the ability of the first and second transducers to exchange the ultrasonic wave signals is not adversely affected by the temperature of the fluid in the pipe. The first and second waveguides may be made of a calcium silicate technical ceramic.

A method of measuring moisture carryover (MCO) in a nuclear reactor includes placing a first gamma detector adjacent to a steam conduit configured to transport steam generated by the core. The method additionally includes detecting a first amount of carryover gamma activity of a first quantity of sodium-24 in the steam within the steam conduit with the first gamma detector. The method also includes detecting a second amount of reference gamma activity of a second quantity of sodium-24 in a reference sample of reactor water from the core with a second gamma detector. The method further includes determining a flow rate of liquid water entrained in the steam based on the first amount of carryover gamma activity detected by the first gamma detector and the second amount of reference gamma activity detected by the second gamma detector.

A method for at least partially disrupting or removing radioactive deposits formed on a surface of a structure in a nuclear water reactor is disclosed. The method includes identifying the structure, taking the structure out of operational service, isolating the structure, contacting the surface of the structure with an aqueous solution, and adding an effective amount of an elemental metal in solid form to the aqueous solution. The effective amount includes an amount to at least partially disrupt or remove the radioactive deposits formed on the surface of the structure. The method is conducted at ambient temperature. The radioactive deposits include oxide-containing radionuclides deposited on the surface. The surface is a primary side structure in the nuclear water reactor.

A system that provides a direct indication of peak fuel rod centerline temperature and peak fuel rod clad temperature than conventionally inferred from the power distribution by directly and continuously measuring the fuel temperatures of the fuel pellets in one or more of the hottest fuel elements in the core. The peak fuel rod clad temperature is then obtained from the maximum measured peak fuel rod centerline temperature in combination with the maximum coolant core exit temperature and the minimum coolant flow rate.

A system that provides a direct indication of peak fuel rod centerline temperature and peak fuel rod clad temperature than conventionally inferred from the power distribution by directly and continuously measuring the fuel temperatures of the fuel pellets in one or more of the hottest fuel elements in the core. The peak fuel rod clad temperature is then obtained from the maximum measured peak fuel rod centerline temperature in combination with the maximum coolant core exit temperature and the minimum coolant flow rate.