935 results about "Nuclear reactor core" 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.

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 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.

An in-core restraint assembly is for a nuclear reactor core including an upper core plate, a lower core support plate and a plurality of fuel assemblies extending longitudinally therebetween. Each fuel assembly includes top and bottom nozzles and a plurality of elongated fuel rods extending therebetween. The in-core restraint assembly includes a first restraint element, such as a spring pack, coupled to the upper core support plate and providing a substantially axial compressive force on the top nozzle of the fuel assembly. An optional second restraint element is structured to be coupled to the lower core plate in order to engage and further restrain the fuel assembly proximate the bottom nozzle. The second restraint element includes a pin member extending from the bottom nozzle of the fuel assembly and received in a socket coupled to the lower core support plate, whereby this mating sustains a longitudinal (vertical) frictional force which must be overcome before fuel assembly lift off can occur.

A method is for establishing a nuclear reactor core loading pattern (LP) for fuel assemblies and burnable absorbers (BAs). The method establishes an optimum LP through the steps of: a) providing nuclear data representing fuel assemblies and BAs in a nuclear reactor core; b) depleting the nuclear data to form a reference core depletion; c) incorporating the nuclear data into a system of linear equations of a nuclear design quality flux solution method; d) defining the system of linear equations to include constraints which accurately represent the neutron physics of the reactor; employing the equations as a constraint matrix for a MIP solver to find an optimum core pattern solution; f) repeating steps b) through e) updating the constraints and objective functions to satisfy specified engineering requirements and establish an optimum core loading pattern. An algorithm for deriving the system of equations is also disclosed.

A method is for establishing a nuclear reactor core loading pattern (LP) for fuel assemblies and burnable absorbers (BAs). The method establishes an optimum LP through the steps of: a) providing nuclear data representing fuel assemblies and BAs in a nuclear reactor core; b) depleting the nuclear data to form a reference core depletion; c) incorporating the nuclear data into a system of linear equations of a nuclear design quality flux solution method; d) defining the system of linear equations to include constraints which accurately represent the neutron physics of the reactor; employing the equations as a constraint matrix for a MIP solver to find an optimum core pattern solution; f) repeating steps b) through e) updating the constraints and objective functions to satisfy specified engineering requirements and establish an optimum core loading pattern. An algorithm for deriving the system of equations is also disclosed.

An in-core neutron monitor that employs vacuum microelectronic devices to configure an in-core instrument thimble assembly that monitors and wirelessly transmits a number of reactor parameters directly from the core of a nuclear reactor without the use of external cabling. The in-core instrument thimble assembly is substantially wholly contained within an instrument guide tube within a nuclear fuel assembly.

The invention aims at providing a diversified engineered safety system for a nuclear reactor. The diversified engineered safety system comprises a loop passive residual heat removal subsystem, a secondary-side passive residual heat removal subsystem of a steam generator fed with water by gravity, a total-pressure active residual heat removal subsystem, a low-pressure passive safety injection subsystem and a water-flooding and air-cooling combined passive containment cooling subsystem. All the subsystems can independently run or can be matched with one other, and parallel and redundant complete residual heat removal channels with diversified principles are built between a reactor core and an ultimate heat sink, so that a main system and the reactor core of the reactor can be quickly and efficiently cooled and the reactor can quickly enter a safety state in the initial stage of an accident. Even if one subsystem has a fault, other subsystems can still complete a complete and effective residual heat removal function for guaranteeing that the reactor is effectively cooled.

A method for determining a fresh fuel bundle design for a core of a nuclear reactor may include defining a plurality of inputs including user-defined target conditions for evaluating one or more reference fresh fuel bundle designs for each of N bundle groups, and generating a response surface based on making single rod-type changes in each (i,j) rod location of each bundle of a given reference bundle design being evaluated for each of the N bundle groups. A search algorithm may be iterated to evaluate a given combination of multiple rod-type changes made simultaneously across each of the N bundle groups using the generated response surface to determine an accepted fuel bundle design for each of the N bundle groups that satisfies the user-defined target conditions.

A laser emitting head for irradiating a portion to be machined with laser beams outputted from a laser unit (624) incorporates an emitting head body having a light guide member (625) in the head for guiding the laser beams; a light converging lens (677) for converging the laser beams transmitted from the light guide member in the head; a reflecting mirror (678) for irradiating the portion to be machined with the converged laser beams; mirror rotating means (684) for holding the reflecting mirror to be rotative around the optical axis of the laser beam; distance-adjustment means (674) for adjusting the relative distance between the reflecting mirror and the converging lens; and moving means (665) for moving the reflecting mirror and the converging lens such that the relative distance is maintained so that the light guide member in the head. The converging lens and the reflecting mirror can be introduced / discharged with respect to a portion to be machined and formed in a narrow gap in an incore structure. As a result, a preventive-maintenance / repair operation using laser beams and arranged to be performed in a narrow gap in the incore structure can safely and efficiently be performed.

A method for determining a fresh fuel bundle design for a core of a nuclear reactor may include defining a plurality of inputs including user-defined target conditions for evaluating one or more reference fresh fuel bundle designs for each of N bundle groups, and generating a response surface based on making single rod-type changes in each (i,j) rod location of each bundle of a given reference bundle design being evaluated for each of the N bundle groups. A search algorithm may be iterated to evaluate a given combination of multiple rod-type changes made simultaneously across each of the N bundle groups using the generated response surface to determine an accepted fuel bundle design for each of the N bundle groups that satisfies the user-defined target conditions.

Example embodiments are directed to methods of producing desired isotopes in commercial nuclear reactors and associated apparatuses using instrumentation tubes conventionally found in nuclear reactor vessels to expose irradiation targets to neutron flux found in the operating nuclear reactor. Example embodiments include assemblies for retention and producing radioisotopes in nuclear reactors and instrumentation tubes thereof. Example embodiments include one or more retention assemblies that contain one or more irradiation targets and are useable with example delivery systems that permit delivery of irradiation targets. Example embodiments may be sized, shaped, fabricated, and otherwise configured to successfully move through example delivery systems and conventional instrumentation tubes while containing irradiation targets and desired isotopes produced therefrom.

The invention relates to the nuclear reactor core power monitoring field and discloses a method for online monitoring the neutron flux distribution of the reactor core. The method bases on M1 internal neutron detectors and M2 external neutron detectors which are arranged on the reactor. According to a reference reactor core model of a renewal reactor that MPhi is equal to (1 / k) FPhi, higher order harmonics are solved. Then, reading data of the internal neutron detectors and the external neutron detectors are combined. The neutron flux distribution in the core of a real reactor is reconstructed online.

The invention discloses an experiment device for simulating retention of in-pile melts obtained by melting of a reactor core of a nuclear reactor and an experiment method. The device consists of a copper heating section, a quartz glass component and a data measurement system. According to the experiment device, a melt hard shell of the reactor core is simulated through a semispherical copper ball shell; decay heat is simulated through a spiral heating wire embedded into the semispherical shell; the inner wall surface of a lower enclosure of a pressure container is simulated through quartz glass, and therefore, a flowing heat exchange phenomenon of a narrow gap between the inner wall surface of the pressure container of the reactor and the melt hard shell can be simulated, and the influence on the retention of the in-pile melts caused by flowing of water in the narrow gap can be researched. The experiment device has the advantages of visible experiment process, high experiment precision, high safety, easiness in operation and the like. A running practice shows that the experiment device runs stably and can thoroughly meet a design requirement.

A metal particulate fuel system is described. The metal fuel system may include particulate metal fuel for use in nuclear reactors. The particulate metal fuel may include a plurality of particles of at least one enriched alloy where the particles are compacted into a fuel column. The metal particulate fuel system may also include a cladding and / or a gas-filled plenum.

The present invention relates to a fast reactor type coupling nuclear reaction implementation method and a nuclear reactor for same. The main contents comprise: a fast reactor type coupling nuclear reaction implementation method, a reactor modular design approach, a fast reactor type coupling nuclear reactor, a reactor core, a fuel element, a nuclear control system, and a proliferation fuel system. The fast reactor type coupling nuclear reactor mainly combusts thorium and nuclear waste, and has inherent security. The reactor main container is composed of a fission pool and a moderating pool that are completely isolated from each other but coupling to each other. A primary coolant is separated from a moderator. A thermal insulation layer is disposed between the fission pool and the moderating pool so that both can perform neutron exchange but heat exchange is blocked. Fast neutrons produced by the fission pool and moderated neutrons reflected by the moderating pool may enable the reactor core to simultaneously perform coupling nuclear reaction of the two types of neutrons. The moderating pool may be provided with the nuclear control system, and ex-core coupling core control may be implemented. The moderating pool is provided with a thorium purification fuel system, and on-line extraction of the purification fuel can be performed, and separation of nuclide is safe and simple, thereby providing a solution to the technical bottleneck of "thorium reactor".

The present invention is to provide a reactor core that allows a nuclear plant to continuously operate for a long term period, for example 15 years or longer, without requiring any fuel exchange, reduces the duration and number of maintenance steps involved in regular plant inspections, markedly improves plant availability and economic efficiency, and is effective in terms of nuclear nonproliferation.A plurality of fuel assemblies 103, themselves obtained by arranging fuel rods 100 and water rods 107 in square lattices, are arranged in a square lattice at a certain pitch. The blades 102a of a cross-shaped (cruciform) control rod 102 in a cross section are inserted into four adjacent spaces formed by four fuel assemblies 100 facing each other. A value of 0.06 cm−1 or greater is selected for the ratio (B / S) of the width (B) of each blade on the cruciform control rod 102 and the surface (S) of the fuel lattice defined by the surface area of a square whose side is equal to the pitch between the fuel assemblies 103.

In a computer-implemented method of designing a nuclear reactor of a given reactor plant, an initial, test reactor core design is generated for the given plant based on a plurality of limits input by a user. The limits include a plurality of transient licensing constraints to be satisfied for operating the given plant. The method includes selecting, from a set of automated tools, one or more automated tools to evaluate the test core design, and operating one of more of selected automated tools to output data for display to the user. The displayed data is related to a core design that satisfies the limits inclusive of the transient licensing constraints.

The invention discloses a nuclear reactor core capable of improving neutron flux rate; the core is a circular structure; the center of the core is a Be thermal neutron well region, the circumference of which is a fuel component region, the circumference of which is a Be reflecting layer; the whole core is a circular structure. The maximum thermal neutron flux rate of the central Be thermal neutron well region can be over twice larger than that of the fuel region ; as the thermal neutron well region can be provided with fuel component with high uranium content, the fuel component center can form fast neutron region, therefore, the generated fission neutron has high flux rate.

A fuel element for nuclear power generation made up of at least one coated UO2 fuel kernel embedded in a matrix containing Zr. The kernel and the element each having a coating that provides desired spacing and fission product barrier protection. The fuel enables the functioning and deployment of small scale nuclear reactors with a various features desired for deployment of such devices in developing nations, with limited electrical distribution infrastructure.

The invention relates to a method for determining (30) the volumetric power distribution of the core of a nuclear reactor using a set of detectors for measuring a neutronic flow provided outside the reactor vessel, and a set of probes for measuring the temperature of the coolant at the outlet of the fuel assemblies. The method (30) comprises the step of determining a first volumetric power distribution using a neutronic calculation code (40) instantaneously solving the diffusion equation and updating the isotopic balance of the core during the fuel depletion based on values of the core operation parameters, and the step of determining a new volumetric power distribution by adjusting (60 and 90) the first volumetric power distribution using the measures provided by the sensors for measuringthe neutronic flow and provided outside the reactor vessel (80) and by the temperature measuring probes (100).

The invention provides a system for monitoring loosening parts of a nuclear reactor and a coolant system, solving the problems of reliability, compatibility, stability and maintenance convenience andcomprising an accelerometer, a charge amplifier, signal conditioning equipment, loosening part detection equipment, alarm processing equipment, sound monitor equipment, data collection equipment, a display, loosening part monitor software and the like. The system adopts a PXI bus technology so as to enable every two modules to be synchronized; a signal conditioning module adopts an FPGA programmable technology; a DSP digital technology is adopted to discriminate a loosening part event; loosening part signal length effective value floating discrimination, time delay, channel recheck combination discrimination, alarm prohibition caused by alarm channel recheck, control rod motion or coolant low pressure, and the like ensure that the system has a strong wrong alarming resistance capacity; the data collection equipment collects data to replace tape records so as to ensure the integrity of event signal records; and event signal waveform display and data storage can carry out display back and secondary analysis on signals.

All possible loading patterns for a nuclear reactor core are searched and optimized for compliance with design constraints. The fuel inventory is divided into a few batches according to coarse levels of reactivity. A recursive enumeration process identifies patterns meeting selected core position constraints, which can be user modified to adjust the search space size. For the batch loading patterns satisfying the constraints, the batches are divided into several smaller batches. A sensitivity matrix linearizing the relationship between fuel assembly position and the depletion model is processed through mixed integer linear programming with branching and bounding to identify an optimal daughter loading pattern. The process is repeated through several levels of batch refinement and selection of optimal daughter patterns, including a level where burnable absorbers are assigned to feed assemblies, until the individual fuel assembly level is reached. The multiple optimal patterns remaining provide a range of solutions.

The invention belongs to the technical field of the nuclear reactor engineering, particularly to the water-cooling two-region multiplication nuclear reactor core and the nuclear reactor adopting the same, wherein the water-cooling two-region multiplication nuclear reactor core includes a fast neutron energy spectrum area adopting the coating grain sheath fuel assembly and a thermal neutron energy spectrum area adopting the dense rod bundle fuel assembly or the coating grain sheath fuel assembly. The water-cooling two-region multiplication nuclear reactor core of the invention adopts the coating grain sheath fuel assembly, which breaks through the restrain of the stainless steel shell to the outlet scream temperature, realizes the higher coolant outlet temperature, meanwhile the fast neutron core is leaked to the thermal neutron energy spectrum core constituted by the rod bundle fuel assembly dense boom at the subcritical to be multiplied.

A method of calibrating excore detectors for a pressurized water reactor (PWR) (1) includes: measuring peripheral core flux signals using excore detectors (33) disposed at a plurality of locations spaced about the periphery of the core (9), and using the measured power distribution from either a core monitoring system (43) or in-core flux measurement (69). Calibration of the excore detectors (33) is broken into two parts: (1) the relation between the excore detector signal and weighted peripheral assembly axial offset, and (2) the relation between weighted peripheral assembly axial offset and core average axial offset. Relation (2) can be determined by a representative neutronics model. Accuracy of the neutronics solution is improved by applying (83) nodal calibration factors, which represent the ratio of the measured three-dimensional power distribution (75) to the nodal predicted three- dimensional power distribution and correct the neutronic results to match what would be measured if predictive scenarios were actually performed in the actual reactor core (9).

The invention belongs to the technical field of nuclear reactors and discloses a heating pipe-type dual-mode space nuclear reactor core. The reactor core comprises a reactor core active zone, a reactor core cylinder, a radial reflection layer, an axial reflection layer and a control drum. The reactor core active zone is located in the reactor core cylinder. The axial reflection layer is located above the reactor core active zone. The axial reflection layer is a hollow cylindrical structure. The reactor core active zone, the axial reflection layer and a heating pipe are located in the cavity ofthe radial reflection layer. The reactor core can effectively prevent the heat pipe overheating and has the beneficial effects of safety and reliability such as passive and non-single point failure.

The invention refers to a filter (1) for separating particle from cooling water in a nuclear plant, and a fuel assembly with such a filter. The filter has an inlet end (2) and an outlet end (3) and permits through-flow of the cooling water in a main flow direction (x). The filter includes a number of sheets (4) extending in the flow direction from the inlet end to the outlet end. The sheets are arranged beside each other and form passages for the cooling water. The sheets include a first portion (4′) extending from the inlet end (2), a second portion (4″) extending from the outlet end (3), and a third portion (4′″) extending between the first portion (4′) and the second portion (4″). The sheets (4) have along the first portion continuous wave-shape extending in a direction (y) transversally to the flow direction (x) and along the third portion a continuous wave-shape extending in the flow direction (x).

The invention relates to an active and passive nuclear steam supplying system based on 177 reactor core. The nuclear steam supplying system comprises a nuclear reactor core and a reactor coolant system, and is characterized in that the nuclear reactor core comprises 177 nuclear fuel assemblies with the active length of 12-14 feet; and the reactor coolant system comprises a reactor pressure vessel, a main pipe connecting an inlet and an outlet of a reactor coolant, a main pump, a steam generator, a pressure stabilizer and a pressure relief tank. The invention also relates to a nuclear power station employing the above active and passive nuclear steam supplying system based on the 177 reactor core. The power of the assembling unit of the nuclear power station is 1000-1400 MWe, the average availability is more than or equal to 90%, the maximum ground acceleration is 0.3 g, and a containment has a double-layer steel-made structure for resisting bump of large business airplanes. The active and passive nuclear steam supplying system based on the 177 reactor core has the functions of alleviating and preventing severe accidents, and a core measuring meter penetrates into the reactor pressure vessel from top to bottom, so that the active and passive nuclear steam supplying system possesses an active manner and passive manner combined residual heat removal system and a digitalizer control diversity protection system.

A method and arrangement of determining pin enrichments for a fuel bundle of a nuclear reactor, where a plurality of input parameters and target conditions may be input and enrichment changes, to be made across the fuel bundle, may be calculated using response matrix technology. Fuel bundle pin enrichment data may be output that satisfies the target conditions. The method and arrangement may enable production of fuel bundles having a desired local peaking, exposure peaking and R-factor performance. Consequently, given fuel cycles typically may be loaded and operated such that less fuel may be needed for identical cycle lengths, potentially resulting in improved fuel cycle economics. Additionally, because fuel bundle development may require fewer iterations, there may be a substantial cycle time reduction in the bundle design process, potentially reducing cost and enhancing profitability.

A heat treatment method for desensitizing a nickel-based alloy with respect to environmentally-assisted cracking, the alloy having the following composition in percentages by weight: C≦0.10%; Mn≦0.5%; Si≦0.5%; P≦0.015%; S≦0.015%; Ni≧40%; Cr=12%-40%; Co≦10%; Al≦5%; Mo=0.1%-15%; Ti≦5%; B≦0.01%; Cu≦5%; W=0.1%-15%; Nb=0-10%; Ta≦10%; the balance being Fe, and inevitable impurities that result from processing, characterized in that the alloy is held at 950° C.-1160° C. in an atmosphere of pure hydrogen or containing at least 100 ppm of hydrogen mixed with an inert gas. A part made of a nickel-based alloy having the composition and that has been subjected to the heat treatment.