9128results about "Nuclear energy generation" patented technology

A control rod drive mechanism (CRDM) for use in a nuclear reactor, the CRDM comprising: a connecting rod connected with at least one control rod; a lead screw; a drive mechanism configured to linearly translate the lead screw; an electromagnet coil assembly; and a latching assembly that latches the connecting rod to the lead screw responsive to energizing the electromagnet coil assembly and unlatches the connecting rod from the lead screw responsive to deenergizing the electromagnet coil assembly. The latching assembly is secured with and linearly translates with the lead screw, while the electromagnet coil assembly does not move with the lead screw. The electromagnet coil assembly is at least coextensive with a linear translation stroke over which the drive mechanism is configured to linearly translate the lead screw.

In a method of producing isotopes in a light water power reactor, one or more targets within the reactor may be irradiated under a neutron flux to produce one or more isotopes. The targets may be assembled into one or more fuel bundles that are to be loaded in a core of the reactor at a given outage. Power operations in the reactor irradiate the fuel bundles so as to generate desired isotopes, such as one or more radioisotopes at a desired specific activity or stable isotopes at a desired concentration.

A control rod drive mechanism (CRDM) comprises a lead screw, a motor threadedly coupled with the lead screw to linearly drive the lead screw in an insertion direction or an opposite withdrawal direction, a latch assembly secured with the lead screw and configured to (i) latch to a connecting rod and to (ii) unlatch from the connecting rod, the connecting rod being free to move in the insertion direction when unlatched, and a release mechanism configured to selectively unlatch the latch assembly from the connecting rod.

A rod assembly for a fuel bundle of a nuclear reactor may include an upper end piece, lower end piece and a plurality of rod segments attached between the upper and lower end pieces and to each other so as to form an axial length of the rod assembly. The rod assembly may include an adaptor subassembly provided at given connection points for connecting adjacent rod segments or a given rod segment with one of the upper and lower end pieces. The connection points along the axial length of the rod assembly may be located where the rod assembly contacts a spacer in the fuel bundle. One (or more) of the rod segments may include an irradiation target therein for producing a desired isotope when a fuel bundle containing one (or more) rod assemblies is irradiated in a core of the reactor.

A sensor device and method of employment is provided. More specifically, a sensor device adapted to detect, identify and/or measure a chemical and/or physical characteristic upon placement of the device into an environment, especially a liquid medium for which monitoring is sought is provided.

Described is a radiation dosimeter including multiple sensor devices (including one or more passive integrating electronic radiation sensor, a MEMS accelerometers, a wireless transmitters and, optionally, a GPS, a thermistor, or other chemical, biological or EMF sensors) and a computer program for the simultaneous detection and wireless transmission of ionizing radiation, motion and global position for use in occupational and environmental dosimetry. The described dosimeter utilizes new processes and algorithms to create a self-contained, passive, integrating dosimeter. Furthermore, disclosed embodiments provide the use of MEMS and nanotechnology manufacturing techniques to encapsulate individual ionizing radiation sensor elements within a radiation attenuating material that provides a “filtration bubble” around the sensor element, the use of multiple attenuating materials (filters) around multiple sensor elements, and the use of a software algorithm to discriminate between different types of ionizing radiation and different radiation energy.

A set of alloy formulations is disclosed to use with thorium based nuclear fuels in a fast spectrum reactor; with thorium based nuclear fuels in existing thermal spectrum power reactors; for medical isotope production in the epithermal, the fast, the fission spectrum and the thermal spectra; and to use as fuel in test and experimental reactors that are non proliferative. The alloys form inert metal matrixes to hold fine particles of dispersed thorium containing fuel. The formulations also are useful for the production of medical and commercial isotopes in the high energy, fast and epithermal neutron spectra.

Systems and methods suitable for capable of autonomous crack detection in surfaces by analyzing video of the surface. The systems and methods include the capability to produce a video of the surfaces, the capability to analyze individual frames of the video to obtain surface texture feature data for areas of the surfaces depicted in each of the individual frames, the capability to analyze the surface texture feature data to detect surface texture features in the areas of the surfaces depicted in each of the individual frames, the capability of tracking the motion of the detected surface texture features in the individual frames to produce tracking data, and the capability of using the tracking data to filter non-crack surface texture features from the detected surface texture features in the individual frames.

A containment vessel includes a primary containment vessel containing a reactor pressure vessel, an upper secondary containment vessel arranged above the primary containment vessel, and a gas-phase vent pipe linking the primary containment vessel and the upper secondary containment vessel by way of an isolation and connection switching system. The gas-phase vent pipe may be arranged either inside or outside the primary containment vessel and the upper secondary containment vessel. Alternatively, it may be embedded in the wall. An igniter may be arranged in the upper secondary containment vessel. The air in the upper secondary containment vessel may be replaced by nitrogen. A gravity-driven flooding system pool may be arranged in the upper secondary containment vessel and cooling water may be led from the inside of the pool to the inside of the primary containment vessel.

A new transmutation assembly permits an efficient transmutation of a long-lived radioactive material (long-lived FP nuclides such as technetium-99 or iodine-129) which was produced in the nuclear reactor. Wire-type members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived FP nuclides are surrounded by a moderator material and installed in cladding tubes to form FP pins. The FP pins, and nothing else, are housed in a wrapper tube to form a transmutation assembly. The wire-type members can be replaced by thin ring-type members. The transmutation assemblies can be selectively and at least partly loaded into a core region, a blanket region or a shield region of a reactor core in a fast reactor. From a viewpoint of reducing the influence on the reactor core characteristics, it is optimal to load the transmutation assemblies into the blanket region.

A method and apparatus for formation testing is disclosed. In a preferred embodiment, a formation testing tool includes a longitudinal body with a flowbore; a plurality of extendable centralizing pistons coupled to the body; an extendable sample device coupled to the body; and a centralizing hydraulic circuit configured to cause each of the plurality of centralizing pistons to extend at substantially the same rate. The centralizing hydraulic circuit includes a series of flow control and pressure-determining valves configured to extend the centralizing pistons at substantially the same rate, and to help maintain stability in the hydraulic circuit in response to external pressures. In some embodiments, the extendable sample device is preferably configured to be recessed beneath a surface of the body in a first position and to extend beyond the surface in a second position. The extendable sample device is preferably extended to contact the borehole wall substantially normal to the wall, protecting the sample device from excessive bending moments and other excessive forces.

A molten salt breeder reactor that has fuel conduit surrounded by a fertile blanket. The fuel salt conduit has an elongated core section that allows for the generation of electrical power on a scale comparable to commercially available nuclear reactors. The geometry of the fuel conduit is such that sub-critical conditions exist near the input and output of the fuel salt conduit and the fertile blanket surrounds the input and output of the fuel salt conduit, thereby minimizing neutron losses.

Provided are an underwater processing apparatus which can effectively prevent water from entering a shield for a workpiece having a surface ruggedness, and in which variation in a gas flow for a processing part is reduced, a processing method and an application thereof to a nuclear reactor, and the under water processing device is composed of a shield means which locally cover the processing part with the gas in order to prevent water from entering the shield member, the shield means having a solid wall formed of a member which is slidable in a part where it make contact with the workpiece, and adapted to make contact with the workpiece and to be moved up and down by a pressing force, and a water jetting means for forming a water curtain around the outer periphery of the solid wall.

A power module assembly includes a reactor core immersed in a coolant and a reactor vessel housing the coolant and the reactor core. An internal dry containment vessel submerged in liquid substantially surrounds the reactor vessel in a gaseous environment. During an over-pressurization event the reactor vessel is configured to release the coolant into the containment vessel and remove a decay heat of the reactor core through condensation of the coolant on an inner surface of the containment vessel.

Textured surface for increasing Leidenfrost temperature. The texture comprises of surface features over multiple length scales—from micro to nanoscale—wherein the features at each length scale have a size, aspect ratio, and spacing selected to increase the Leidenfrost temperature. The structure includes an array of microscale structures disposed on the surface, the structure having size, aspect ratio and spacing selected to increase Leidenfrost temperature. The microscale structures may also include nanoscale structures on their surface to create a hierarchical structure. The structures result in an increased Leidenfrost temperature.

The invention is an integrated temporary wireless network system for controlling, collecting, processing, and responding to data generated by a plurality of sensors configured for measuring data associated with operation of, and the environment related to, commercial, industrial and manufacturing operations, processes and equipment located within a fixed geographic area; a plurality of Wireless-Nodes, one coupled to each sensor, configured and adapted for wirelessly transmitting the data, via a Repeater if necessary, to a Base-Station, configured for receiving the data and connected to a primary computer; a CPU coupled with the primary computer and; a memory operatively connected to the CPU, the memory containing a program adapted to be executed by the CPU and the CPU and memory cooperatively adapted to perform data flow control from the plurality of Wireless-Nodes, data storage, and diagnostic analysis of the transmitted sensor data.

The invention belongs to safety equipment for a nuclear power plant, in particular to a passive and active combined special safety system for the nuclear power plant. The system comprises a passive high-pressure core water supply tank, an accumulator, a low-pressure safety injection pump, a refueling water tank, a secondary-side passive residual heat removal heat exchanger, a vapor condensation water tank, a containment spray system, a passive reactor cavity water injection system, related valves and related pipes. When the nuclear power plant has a design basis accident or beyond design basis accident, through a series of passive and active safety facilities used in all steps, the circuit of a reactor and the core can be cooled effectively, so that the nuclear power plant can enter a safe cold shutdown condition smoothly; and thus, the serious accident consequence of the reactor is inhibited or relieved, damage caused by accidents is relieved, and the safety of the nuclear power plant is improved.

A power module assembly includes a reactor vessel containing a reactor core surrounded by a primary coolant. A containment vessel is adapted to be submerged in a containment cooling pool and to prohibit a release of the primary coolant outside of the containment vessel. A secondary cooling system is configured to remove heat generated by the reactor core. The heat is removed by circulating liquid from the containment cooling pool through the primary coolant.

A method for producing ultra-pure Pu-238 is provided. The method comprises the steps of short-term irradiating Am-241 targets with a high, thermal neutron flux greater than 6.5×10¹⁴ neutrons cm⁻² s⁻¹ and more preferably greater than 1×10¹⁵ neutrons cm⁻² s⁻¹ for a predetermined period of time preferably from 20 days to 30 days and more preferably approximately 25 days to convert a substantial fraction of the Am-241 to Cm-242, e.g., approximately 0.555 g of Cm-242 per g of Am-241 charged, thereafter promptly chemically separating the produced Cm-242, preferably within 10 days to 20 days of the irradiation cycle of the Am-241 targets and more preferably within about 15 days of the irradiation cycle of the Am-241 targets, and recovering the ultra-pure Pu-238 decay product of the separated Cm-242.

A system and method for displaying information related to the status of at least one machine of a plurality of machines. This system includes a data module coupled to each of the plurality of machines. The data module is adapted to collect and store information related to the corresponding machine. A data link is coupled to the data module. A display module is coupled through the data link. The data link is adapted to transmit data information between the data module coupled to each machine and the display module. The display module is adapted to display information relating from only a subset of the plurality of machines. The subset is selected from the plurality of machines as a function of machine parameters input by a user.

The design of a compact, high-efficiency, high-flux capable compact-accelerator fusion neutron generator (FNG) is discussed. FNG's can be used in a variety of industrial analysis applications to replace the use of radioisotopes which pose higher risks to both the end user and national security. High efficiency, long lifetime, and high power-handling capability are achieved though innovative target materials and ion source technology. The device can be scaled up for neutron radiography applications, or down for borehole analysis or other compact applications. Advanced technologies such as custom neutron output energy spectrum, pulsing, and associated particle imaging can be incorporated.

The invention belongs to a containment cooling system, in particular, to a high-capacity and fully passive containment cooling system. The containment cooling system is used for collecting thermotechnical parameters related to the containment by using a sensor and tracking a cooling process; a plurality of storage boxes for storing different coolants is arranged above the containment; the dynamic control of containment cooling power can be realized by choosing the types of the coolants and adjusting the flows of the coolants; and by using a generalized passive control unit, the start-up and the whole running process of the containment cooling system can be completely independent on external power supply, thus the containment cooling system has a characteristic of fully passiveness.

A support grid for a nuclear fuel assembly, the fuel rod assembly having a generally cylindrical fuel rod with a diameter, wherein the support grid includes a frame assembly having a plurality of generally uniform cells, each the cell having at least one wall and a width and at least one generally cylindrical tubular member having a cell contact portion with a greater diameter and at least one helical fuel rod contact portion with a lesser diameter, the cell contact portion and the fuel rod contact portion joined by a transition portion, the greater diameter being generally equivalent to the cell width, and the lesser diameter being generally equivalent to the fuel rod diameter such that a fuel rod disposed in the tubular member would engage the inner diameter. Wherein the least one tubular member disposed in one cell of the plurality of generally square cells so that the cell contact portion engages the at least one cell sidewall.

A fusion reactor apparatus for initiating a fusion reaction in a fusionable material is disclosed. The apparatus includes a vessel operable to contain a liquid medium and a vortex generator operable to generate a vortex in the liquid medium. The apparatus also includes a plasma generator operable to generate a magnetized plasma of the fusionable material and to introduce the magnetized plasma into the vortex and a pressure wave generator operably configured to cause a pressure wavefront in the liquid medium to envelope the magnetized plasma and to converge on the magnetized plasma to impart sufficient energy to the fusionable material to initiate fusion in the fusionable material.

Embodiments of systems and methods for compressing plasma are described in which plasma pressures above the breaking point of solid material can be achieved by injecting a plasma into a funnel of liquid metal in which the plasma is compressed and/or heated.

A laser maintenance apparatus is provided with a laser system including a laser light source and an optical system and emitting a laser light, a light transmitting device for generating the laser light emitted from the laser system, and a laser irradiation device for irradiating the laser light generated by the light transmitting device to an object portion. The laser system includes an element for changing an irradiation condition of the laser light.

Provided are methods and devices for compression of a spheromak plasma (e.g., deuterium-trirtium (D-T)-derived) in a magnetic well configured within a plasma combustion chamber using an induction coil axially adjacent to the plasma, wherein a moveable member (e.g., piston, cam and follower) drives the induction coil toward the plasma, pushing the plasma via magnetic pressure into the magnetic well and compressing the plasma substantially adiabatically (e.g., coil motion is well below the plasma sound speed). The compression quickly increases both plasma density and temperature past the point of ignition, and after plasma burn, the coil is backed-off to allow the plasma to re-expand, providing for refueling and repetition of the compression cycle. Additionally provided are spaced annular plasma formation electrodes, suitably configured for generating and injecting magnetized plasma into a plasma combustion chamber. Preferably, spaced annular plasma formation electrodes are used in combination with moveable compression members as disclosed herein.

Improved methods for the extraction or dissolution of metals, metalloids or their oxides, especially lanthanides, actinides, uranium or their oxides, into supercritical solvents containing an extractant are disclosed. The disclosed embodiments specifically include enhancing the extraction or dissolution efficiency with ultrasound. The present methods allow the direct, efficient dissolution of UO2 or other uranium oxides without generating any waste stream or by-products.

A fuel assembly spacer grid including swirl deflectors respectively arranged at interconnections between interconnecting longitudinally and laterally-extending straps on upper ends of the interconnecting straps adapted to support fuel elements of a nuclear fuel assembly. Each of the swirl deflectors has four vanes bent to have an air vane shape. By virtue of this configuration, an improvement in the efficiency of the spacer grid cooling the fuel elements. The spacer grid also includes springs each configured to generate not only a main spring force caused by a displacement of the spring occurring when the spring comes into contact with a fuel element placed in a reactor core, but also an additional spring force caused by hydraulic pressure applied to the spring. Each spring, which is in a fixed state at one end thereof, has a free bent portion at the other end. By virtue of such a spring configuration, it is possible to compensate for a reduction in the initial spring force of the spring resulting from a change in the property of the spring material. The spring has a curved contact portion configured in such a manner that it is in conformal surface contact with a circumferential surface of the fuel element, thereby enhancing vibration suppressing and abrasion resistance forces.

The invention relates to an active-passive combined reactor core residual heat removal system for a nuclear power station. The active-passive combined reactor core residual heat removal system comprises a safety injection system, a containment spraying system, an auxiliary water supply system, a reactor cavity water injection system, a secondary side passive residual heat removal system, a passive containment heat leading-out system and related valves and pipelines. The active-passive combined reactor core residual heat removal system provided by the invention realizes three safety functions of controlling reactivity, discharging reactor core heat and containing radioactive substance when an accident occurs in an active-passive combined multi-redundancy diversity manner. The active-passive combined reactor core residual heat removal system provided by the invention can completely realize the safety functions of safety injection and safety spraying under the condition that the accident occurs in a nuclear power plant, reactor cavity water injection under the serious accident working condition and the like, effectively improve reliability of a safety system, enhance coping capability of the safety system under the working condition that the accident occurs in the nuclear power station, can effectively prevent and relieve the serious accident, reduce the reactor core melting probability and risk probability of large-scale radioactivity release and greatly improve safety performance of the nuclear power station.