213results about "Steam generator primary side" patented technology

Apparatus and a method to inspect tubing by means of a free flying, autonomous inspection head that is not attached by wires to external control and data acquisition equipment. The inspection head travels through the tube with an attached module that integrates all the necessary support for the electronic and mechanical control of a nondestructive sensor within the inspection head.

A steam generator system for a pressurized water reactor which employs an external to containment steam drum and recirculation loop piping. The steam generator system changes the arrangement of a typical pressurized water reactor recirculation steam generator by relocating the functions of steam separation and feedwater preheating outside of the reactor coolant system. The steam generator system and thermal hydraulic conditions are selected in order to minimize the size of the steam generator heat exchanger component volume inside of the containment. The external steam drum component can be isolated in accident conditions when desired and is used as a source of secondary fluid inventory for improved decay heat removal capability and tolerance for loss of feedwater events. Thus, the steam generator component volume inside of the containment is reduced and the amount of maintenance required for the reactor coolant system components are similarly reduced.

A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant, may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

A nuclear reactor including a reactor core comprising a plurality of fuel materials and a plurality of heat pipes. The nuclear reactor further includes a heat exchanger coupled to the reactor core defining a flow path in an open volume including at least two heat pipes of the plurality of heat pipes. Methods of operating a nuclear reactor include passing fluid through an open volume in a heat exchanger including at least two heat pipes extending from a reactor core.

A pressurized water reactor (PWR) includes a cylindrical pressure vessel defining a sealed volume, a nuclear reactor core disposed in a lower portion of the cylindrical pressure vessel, one or more control rod drive mechanisms (CRDMs) disposed in the cylindrical pressure vessel above the nuclear reactor core, and an annular steam generator surrounding the nuclear reactor core and the CRDM. In some such PWR, a cylindrical riser is disposed coaxially inside the pressure vessel and inside the annular steam generator and surrounds the nuclear reactor core and the CRDM, and the steam generator is disposed coaxially inside the cylindrical pressure vessel in an annular volume defined by the cylindrical pressure vessel and the cylindrical riser. In other such PWR, the steam generator is disposed coaxially outside of and secured with the cylindrical pressure vessel.

A nuclear reactor cooling system with passive cooling capabilities operable during a loss-of-coolant accident (LOCA) without available electric power. The system includes a reactor vessel with nuclear fuel core located in a reactor well. An in-containment water storage tank is fluidly coupled to the reactor well and holds an inventory of cooling water. During a LOCA event, the tank floods the reactor well with water. Eventually, the water heated by decay heat from the reactor vaporizes producing steam. The steam flows to an in-containment heat exchanger and condenses. The condensate is returned to the reactor well in a closed flow loop system in which flow may circulate solely via gravity from changes in phase and density of the water. In one embodiment, the heat exchanger may be an array of heat dissipater ducts mounted on the wall of the inner containment vessel surrounded by a heat sink.

A method of storing heat includes moving a portion of a heated fluid from at least one reactor core to at least one tank having solid media, storing heat from the portion of the heated fluid in the solid media, and transferring the stored heat from the solid media to a fluid that can be used by a power plant to generate electrical energy. A system for storing heat in a nuclear power plant includes at least one tank comprising solid media structured and arranged to store heat and an arrangement structured and arranged to pass a first fluid through the at least one tank, transfer heat from the first fluid to the solid media, store the heat in the solid media, and transfer the heat from the solid media to a second fluid. This Abstract is not intended to define the invention disclosed in the specification, nor intended to limit the scope of the invention in any way.

A sealed stopper for an opening in a tubing for joining a chamber and a piping including a fastening ring is provided. The stopper includes a rigid bearing plate and a sealing member carried by the bearing plate, and includes a seal having a planar and flexible central portion with a reduced thickness extending below the bearing plate and a peripheral portion radially deformable by a central expander of the peripheral portion against the inner surface of the fastening ring.

An improved method of inspecting the tubes of a steam generator of a nuclear reactor involves collecting historic data regarding the tube sheet transition regions of each tube for use during a subsequent analysis to create a new simpler signal from which historic artifacts have been removed.

A nuclear reactor cooled with pressurized water, having a pressurized tank installed in which are compact steam generators; each steam generator comprises a plurality of heat-exchange tubes having respective spiral portions set in levels on top of one another to form at least one annular tube bundle delimiting a substantially cylindrical internal central zone, pre-arranged for supply from above with primary water, which then traverses the tube bundle radially.

The invention discloses a pressurized water reactor nuclear power unit secondary circuit thermodynamic system self-adaptive steam supply system and method. By means of the system and the method, a main steam bypass can be self-adaptively put into use or cut out to supply a heating steam source to a low-pressure heater, a deaerator and a high-pressure heater according to secondary circuit thermodynamic parameters in the unit start-stop and normal running process, the defect of long-time hot stand-by duty of an auxiliary boiler is avoided, energy cascade recycling is achieved, meanwhile the reliability, stability and safety are high, and the power consumption is low.

The invention discloses a miniature steam generator thermal hydraulic analysis testing system, and belongs to the technical field of nuclear power station steam generator equipment. The miniature steam generator thermal hydraulic analysis testing system comprises a first loop unit, a second loop unit, a steam generator and a data acquisition unit. The first loop unit and the second loop unit are connected with the steam generator through pipelines, and fluid in the first loop unit can be used for heating fluid in the second loop unit. The data acquisition unit is used for acquiring the temperature, pressure and flow of fluid of the first loop unit and the second loop unit and flowing into steam generator, the first loop unit heats cooling water by means of a heat regenerator and a heating water tank which replace a nuclear reactor being of a complex structure. With respect to the first loop unit and the second loop unit, heat regenerators are employed to lead in a regenerative cycle, energy is saved, and the energy utilization rate of the system is increased.

A nuclear reactor cooling system with passive cooling capabilities operable during a reactor shutdown event without available electric power. In one embodiment, the system includes a reactor vessel with nuclear fuel core and a steam generator fluidly coupled thereto. Primary coolant circulates in a flow loop between the reactor vessel and steam generator to heat secondary coolant in the steam generator producing steam. The steam flows to a heat exchanger containing an inventory of cooling water in which a submerged tube bundle is immersed. The steam is condensed in the heat exchanger and returned to the steam generator forming a closed flow loop in which the secondary coolant flow is driven by natural gravity via changes in density from the heating and cooling cycles. In other embodiments, the cooling system is configured to extract and cool the primary coolant directly using the submerged tube bundle heat exchanger.

An inspection device for inspecting a steam generator having a first long and flexible video probe intended to be inserted inside the housing of the steam generator and configured to be moveable through a fluid passage orifice defined by a spacer plate and a flow tube, all arranged inside the housing. The first video probe includes at least one ferromagnetic element. The inspection device also includes a second long and flexible probe to be inserted into a flow tube, the second probe including at least one permanent magnet arranged to cooperate with the at least one ferromagnetic element of the first probe so that the first probe can be driven along the exterior surface of a flow tube by the second probe when the second probe is inserted into the flow tube and moved therein, and when the first probe is inserted into the housing of the steam generator and positioned near the exterior surface of the flow tube in which the second probe is inserted.

An inspection assembly for insertion inspection of an elongate tubular member. The inspection assembly includes a probe head with a sensor. The assembly also includes a flexible shaft connected to the probe head and transmitting a motive force to the probe head to move the probe head within the elongate tubular member. The probe head includes at least one characteristic to minimize resistance against movement of the probe head along a torturous path within the tubular member.

A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.

This disclosure describes new high temperature, radiation-resistant, ferritic-martensitic steel compositions. The new steels generally contain 9.0-12.0 wt. % Cr, 0.001-1.0 wt. % Mn, 0.001-2.0 wt. % Mo, 0.001-2.5 wt. % W, and 0.1-0.3 wt. % C, with the balance being primarily Fe. More specifically, steels having from 10.0-12.0 wt. % Cr are considered particularly advantageous. Small amounts of N, Nb, V, Ta, Ti, Zr, and B may or may not also be present, depending on the particular embodiment. Impurities may be present in any embodiment, in particular impurities of less than 0.01 wt. % S, less than 0.04 wt. % P, less than 0.04 wt. % Cu, less than 0.05 wt. % Co, and less than 0.03 wt. % As are contemplated. Examples of these steels exhibit improved fracture toughness and reduced thermal creep and swelling.

A U-tube steam generator having a dual system for collecting loose parts and sludge. A loose parts collector having a water overflow edge is disposed between a feedwater inlet and a tube bundle of the steam generator. A sludge collector having a water outlet is disposed downstream of the overflow edge of the loose parts collector and maintains a pressure differential between a water inlet of the sludge collector and the water outlet.

The invention relates to a flexible tube sheet structure for a nuclear power plant steam generator. The flexible tube sheet structure comprises a thermal expansion buffer section and a flexible tube sheet; the two ends of the flexible tube sheet are connected with the thermal expansion buffer section through a transition section; the thermal expansion buffer section is in a barrel shape in the whole and is composed of an inner barrel, an outer barrel and a connection flange, wherein the inner barrel and the outer barrel are different in thickness, a cavity serving as a fluid dead zone is formed between the inner barrel and the outer barrel, the two ends of the connection flange are welded to one end of the inner barrel and one end of the outer barrel, the other end of the inner barrel is welded to the end of the transition section, and the side wall of the other end of the outer barrel is welded to a steam generator barrel body; the flexible tube sheet and the transition section are subjected to primary forging forming, the transition section is a circular-arc-shaped rotation body, and heat exchange tube holes are uniformly and densely distributed in the flexible tube sheet; and the flexible tube sheet, the transition section and the inner barrel of the thermal expansion buffer section are equal in effective thickness. Through the flexible tube sheet structure, the thickness ofa tube sheet can be decreased, thermal stress of the two sides of the tube sheet is decreased, in addition, the tube sheet has certain flexibility, and the thermal expansion difference between the heat exchange tube and the steam generator barrel body can be decreased to a certain extent.

The invention discloses a thermotechnical design method of a sodium water once-through steam generator. The method comprises the following steps: determining thermal parameters of the sodium water once-through steam generator and geometrical parameters except the length of a heat transfer tube; establishing a steam generator single-tube model, and setting the initial pressure and temperature of the steam generator; obtaining the fluid temperature of each fluid control body by using a power grid method; dividing a heat exchange area to calculate a heat exchange coefficient and a heat exchange amount, and solving a wall temperature according to an energy balance relation of heat conduction and convective heat exchange; solving the length of each fluid control body on the heat transfer pipe according to the heat balance equation and the heat transfer equation; calculating the pressure drop of each fluid control body to obtain new pressure distribution. The invention relates to a universalthermal design calculation method for a sodium-water once-through steam generator. The method is convenient to operate, high in universality, flexible to use and high in precision, the thermal designcalculation requirement of the sodium-water once-through steam generator can be met, and a software basis can be provided for independent design of the sodium-cooled fast reactor in China.

There is provided a Cr-containing austenitic alloy tube, wherein a chromium oxide film with a thickness of 0.05 to 1.5 μm having the relationship defined by Formula (i) is formed on the inner surface of the tube, wherein the average concentration of C in the depth range of 5 to 10 μm from the inner surface is lower than the concentration of C in a base metal;

0.4≦δ1/δ2≦2.5  (i)

wherein δ1 and δ2 are thicknesses (μm) of the chromium oxide film at both ends of tube, respectively.