The present disclosure provides a stopped cooling system including: a steam line connecting portion connected to a steam line so as to receive cooling water through the steam line connected to an outlet of a steam generator; a stopped cooling heat exchanger for receiving cooling water that enters the stopped cooling system through the steam line connecting portion, and discharging same through a passage of the heat exchanger; a stopped cooling pump activated to perform stopped cooling of the nuclear reactor upon normal stoppage of the nuclear reactor after primary cooling of the nuclear reactor cooling system or when an accident occurs, and for forming a circulating flow of cooling water that circulates between the steam generator and the stopped cooling heat exchanger; and a water supplying pipe connecting portion connected to the heat exchanger passage and a water supplying pipe, which is connected to the inlet of the steam generator, so as to supply the cooling water cooled in the stopped cooling heat exchanger to the steam generator through the water supplying pipe.

An integrated nuclear-powered heat pump system includes a nuclear power plant including a nuclear reactor coolant and may be configured to generate electricity. The system additionally includes a heat pump including a refrigerant as a working fluid. The heat pump is integrated with the nuclear power plant so as to be in at least thermal contact with the nuclear reactor coolant. The electricity generated by the nuclear power plant may be used to drive the heat pump. The system is instrumental with regard to generating heat for industrial applications.

The disclosure is directed to a system and techniques for integrating sensors within a generator support to detect fluctuations. Such techniques may be performed by a device that includes at least one fiber optic sensor and may comprise receiving, via the at least one fiber optic sensor, an optical signal transmitted over at least one fiber optic link integrally formed with a conduit support, the conduit support coupled with a steam generator conduit. The techniques may further comprise generating, based on the optical signal, strain data related to the steam generator conduit, and based on the generated strain data, determine one or more oscillatory characteristic of a steam generator associated with the steam generator conduit.

An autonomous self-powered system for cooling radioactive materials comprising: a pool of liquid; a closed-loop fluid circuit comprising a working fluid having a boiling temperature that is less than a boiling temperature of the liquid of the pool, the closed-loop fluid circuit comprising, in operable fluid coupling, an evaporative heat exchanger at least partially immersed in the liquid of the pool, a turbogenerator, and a condenser; one or more forced flow units operably coupled to the closed-loop fluid circuit to induce flow of the working fluid through the closed-loop fluid circuit; and the closed-loop fluid circuit converting thermal energy extracted from the liquid of the pool into electrical energy in accordance with the Rankine Cycle, the electrical energy powering the one or more forced flow units.

A steam generator includes a shroud and an annular stepwise helical baffle extending along at least part of a length of the shroud. There is a riser located in a central region of the steam generator. The helical baffle is made up of at least one annuler sector of flat plates. The edges of the flat plates may be straight or corrugated.

The present invention discloses a heat exchange system and a nuclear reactor system. The heat exchange system includes: a heating device; a heat consuming device connected with the heating device through a pipe to form a loop; and a steam, which is in a wet steam state before being supplied to a heat source, and is supplied to the heat consuming device after becoming dry steam or superheated steam by exchanging heat with the heating device. Heat exchange efficiency and security of the nuclear reactor system are improved by adopting steam as a heat exchange medium.

A neutron absorbing insert for use in a fuel rack. In one aspect, the insert includes: a plate structure having a first wall and a second wall that is non-coplanar to the first wall; the first and second walls being formed by a single panel of a metal matrix composite having neutron absorbing particulate reinforcement that is bent into the non-coplanar arrangement along a crease; and a plurality of spaced-apart holes formed into the single panel along the crease prior to bending.

A steam generator tube probe includes a probe head comprising an electronic probe coupled between a proximal portion of the head that is configured for entry into a steam generator tube and a distal portion of the head; and a probe shaft coupled to the distal portion of the shaft and comprising a flexible metallic conduit that comprises a plurality of interlocking portions, each interlocking portion moveably affixed to at least one adjacent interlocking portion.

A steam boiler liquid separator having a plurality of liquid separator segments connected together to form a shaft disposed inside a pipe section, with swirl vanes disposed around the shaft. The liquid separator segments each include a shaft segment connected to a pipe segment by a swirl vane such that the shaft segment is disposed radially inward from the pipe segment.

Disclosed is an apparatus and method for conditioning fluid flow in a nuclear power plant steam generator. A flow conditioning device includes an outer enclosure defining a plurality of entrance apertures arranged in an array and a plurality of exit apertures arranged in an array. A plurality of baffle plates are defined within the outer housing. The baffle plates define flow channels in fluid communication with the entrance and exit apertures to create a flow path of alternating directions. The flow channels receive fluid flow from the plurality of entrance apertures, direct the fluid flow from the entrance apertures in alternating directions through the flow channels to impart turning and frictional pressure loss to the fluid flow, and direct exiting fluid flow through the exit apertures into the tubelane region of the steam generator.