Abstract

Provided in the present invention are a construction issue evaluation method and system for a nuclear power plant under construction. The construction issue evaluation method for a nuclear power plant under construction comprises: step S1, acquiring a list of construction issues of a nuclear power plant under construction; step S2, performing quantitative evaluation and qualitative evaluation on an issue in the list of construction issues, so as to obtain issue importance levels; and step S3, selecting the more serious one of the issue importance level in the quantitative evaluation and the issue importance level in the qualitative evaluation as the final level of the issue, wherein the quantitative evaluation is performed on the basis of the degree of contribution of a structure, a system and/or a device involved in the issue to the overall risk of a nuclear power plant, and is performed on the basis of the degree of influence of the issue on the function implementation of the structure, the system and/or the device itself. Graded and classified evaluation of an issue is realized, and thus an internal supervision and inspection issue found in a nuclear power plant in a construction stage can be scientifically evaluated, and supervision and inspection work of the nuclear power plant under construction can be effectively supported and carried out.

IPC Classes  ?

• G06Q 10/0637 - Strategic management or analysis, e.g. setting a goal or target of an organisationPlanning actions based on goalsAnalysis or evaluation of effectiveness of goals

Abstract

The present invention provides an integrated reactor safety system and a control method therefor. The system comprises an integrated reactor pressure vessel and a containment building. The integrated reactor pressure vessel is arranged in the containment building, and a primary and secondary side heat exchanger is arranged in the integrated reactor pressure vessel, a secondary side inlet of the primary and secondary side heat exchanger is connected to secondary side water supply piping, and a secondary side outlet of the primary and secondary side heat exchanger is connected to secondary side outlet piping; a passive residual heat discharge heat exchanger is arranged on an outer side of the containment building; an outlet of the passive residual heat discharge heat exchanger is connected to the secondary side water supply piping by means of residual heat discharge outlet piping, and an inlet of the passive residual heat discharge heat exchanger is connected to the secondary side outlet piping by means of a residual heat discharge inlet piping. The present invention meets design-basis accident mitigation requirements for reactors, ensures reactor safety, simplifies the equipment to the maximum extent, improves arrangement space utilization, and is more economical.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A four-active one-hot-standby frequency converter system and a control method therefor. The four-active one-hot-standby frequency converter system comprises four main pump branches, a hot-standby branch and a redundant main unit, wherein each of the four main pump branches comprises a busbar, a main pump and a main pump frequency converter, an input end of the main pump frequency converter being connected to the busbar by means of a main pump feeder circuit breaker, and an output end of the main pump frequency converter sequentially being connected to the main pump by means of an outlet circuit breaker and a main pump circuit breaker; the hot-standby branch comprises a hot-standby frequency converter, an input end of the hot-standby frequency converter being connected to any busbar in the four main pump branches by means of a hot-standby frequency converter feeder circuit breaker, and an output end of the hot-standby frequency converter being connected to all of the four main pump branches by means of an interlocking switching apparatus; and the redundant main unit is used for instructing the hot-standby branch to connect to one of the four main pump branches. When any main pump frequency converter is in a faulty state, a common hot-standby frequency converter can be rapidly switched by means of adding a hot-standby branch and newly adding an outlet circuit breaker to a main pump branch.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• F04B 49/00 - Control of, or safety measures for, machines, pumps, or pumping installations, not otherwise provided for in, or of interest apart from, groups
• F04B 49/06 - Control using electricity
• F04B 49/20 - Control of, or safety measures for, machines, pumps, or pumping installations, not otherwise provided for in, or of interest apart from, groups by changing the driving speed

Abstract

A passive nuclear steam supply system (100), comprising a reactor body (1), the reactor body (1) comprising a reactor pressure vessel (11); and a reactor coolant system (2), the reactor coolant system (2) comprising two coolant loops (21), a pressurizer (22) and a surge line (23), and each coolant loop (21) comprising a steam generator (24), two primary pumps (25), a hot leg main pipe (26) and two cold leg main pipes (27), wherein the hot leg main pipe (26) is provided with a first liquid level pressure tap (261) and a second liquid level pressure tap (262), the first liquid level pressure tap (261) being located at the bottom of the hot leg main pipe (26), and the second liquid level pressure tap (262) approaching the steam generator (262) and being located at the top of the hot leg main pipe (26).

IPC Classes  ?

• G21C 15/14 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts conducting a hot fluidArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from ducts comprising auxiliary apparatus, e.g. pumps, cameras

Abstract

A passive reactor core protection mechanism based on siphon breakage, and a reactor cooling system. The passive reactor core protection mechanism based on siphon breakage comprises: a heat exchanger (4); a circulation pump (5), an outlet end of the circulation pump (5) being in communication with an inlet end of the heat exchanger (4); an inlet pipeline (1) with one end being inserted into a reactor core storage pool (3) and the other end being in communication with an outlet end of the heat exchanger (4); an outlet pipeline (2) with one end being in communication with an outlet end of a reactor core (7) and the other end being in communication with an inlet end of the circulation pump (5); and a siphon breakage mechanism (6) arranged in the inlet pipeline (1) and/or the outlet pipeline (2) and configured to break a siphon effect caused by a rupture in the inlet pipeline (1) and/or the outlet pipeline (2).

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A system for producing radioactive isotopes by using a heavy water reactor nuclear power plant, comprising: a production channel (2) comprising a straight pipe section (211), a bent pipe section (212) and an inclined pipe section (213) which are connected together in sequence, wherein the bottom of the straight pipe section (212) is located at the bottom of the inner side of a heavy water reactor calandria vessel (8), the inclined pipe section (213) is provided with a charging port (21) and a discharging port (22), and the charging port (21) is closer to the bent pipe section (212) than the discharging port (22); a target box carrier (3) provided in the production channel (2) and used for bearing a target box (4) and driving the target box (4) to move in the production channel (2); a target box traction mechanism (1) provided at the tail end of the inclined pipe section (213) away from the bent pipe section (212) and used for pulling the target box carrier (3); an automatic transport mechanism provided above the calandria vessel (8) and used for transporting the target box (4) before irradiation production and receiving the irradiated target box (4) in the production channel (2); and a conveying mechanism (10) used for conveying the target box (4) before irradiation production to the charging port (21) and placing the target box (4) in the target box carrier (3). Mutual impact of the target box (4) is avoided, radiation borne by the target box traction mechanism (1) is reduced, and disturbance to a reactor core is reduced.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors

Abstract

The present invention relates to a site boundary atmospheric dispersion factor analysis method and system for a small nuclear reactor. The method comprises: when the distances from a reactor in all orientations are all greater than a preset distance, using a first calculation method to calculate atmospheric dispersion factor values in all orientations within a plurality of standard time periods; and when none of the distances from the reactor in one or more orientations is greater than the preset distance, if the release mode of a release source is selected to be ground-level release, using a second calculation method to calculate atmospheric dispersion factor values in all orientations within the plurality of standard time periods, and if the release mode of the release source is selected to be elevated release, using the first calculation method and the second calculation method to respectively calculate atmospheric dispersion factor values in all orientations within the plurality of standard time periods. Calculation using the two calculation methods in different cases effectively solves the problem of doses for personnel at site boundaries being close to limit values in current small nuclear reactor analysis, etc.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation

Abstract

The present invention relates to an apparatus and method for measuring acoustic excitation noise in a steam pipeline of a nuclear power plant. Multiple layers of sound absorption materials are arranged at the end in a shielding cavity close to a bottom side, and a flexible attaching mechanism is arranged at an opening of the shielding cavity; during measurement, the flexible attaching mechanism is attached to the outer wall of a steam pipeline of an nuclear power plant; the multiple layers of sound absorption materials isolate noise in an environment outside the shielding cavity; by means of a through hole, a microphone is placed in the shielding cavity for measurement, implementing non-destructive indirect measurement of noise in the steam pipeline of the nuclear power plant from the outside of the steam pipeline, and realizing the advantages of interference prevention and easy installation; the noise isolation of the shielding cavity and the multiple layers of sound absorption materials and the tight attachment of the flexible attaching mechanism on the pipe wall eliminate the impact of field environments on measurement; and the direct attachment of the shielding cavity to the pipe wall solves the problem that measurement cannot be carried out due to a narrow space size, realizing effective noise tests in high-temperature and high-pressure environments.

IPC Classes  ?

• G01H 17/00 - Measuring mechanical vibrations or ultrasonic, sonic or infrasonic waves, not provided for in the other groups of this subclass

Abstract

A universal pump casing (1) for a nuclear power main pump, the universal pump casing comprising: a spherical housing (11); a discharge section (12), which is arranged on one side of the spherical housing (11) and configured to connect to a cold-end pipeline of a system; a suction section (14), which is arranged at an axial upper end of the spherical housing (11) and configured to connect to a steam generator (2) and a suction guide pipe (3) of a main pump; and an annular end cover (15), which is located at an axial lower end of the spherical housing (11) and configured to connect to a main pump cartridge (6), wherein the annular end cover (15) is provided with a plurality of main bolt holes (151), which are spaced apart; and one side of each main bolt hole (151) is provided with a positioning spigot (152), and the other side thereof is provided with a positioning end face (153).

IPC Classes  ?

• F04D 29/42 - CasingsConnections for working fluid for radial or helico-centrifugal pumps
• F04D 29/44 - Fluid-guiding means, e.g. diffusers

Abstract

The present invention provides a temperature field analysis method and system for natural air cooling of a control rod drive mechanism. The analysis method comprises: establishing a control rod drive mechanism model, and setting different external air temperatures to perform simulation to obtain corresponding surface heat fluxes and first surface temperatures of the control rod drive mechanism model under different thermal loads; establishing a natural air cooling model, and performing simulation on the basis of the surface heat fluxes to obtain second surface temperatures of the control rod drive mechanism model; and on the basis of an interpolation method, determining a balance point temperature between the first surface temperatures and the second surface temperatures, wherein the control rod drive mechanism model comprises an internal flow channel, and the natural air cooling model is an integrated head package natural air cooling model. On the basis of the principles of sensitivity analysis, thermosyphon of a single drive mechanism model and natural air cooling of a full-field model are subjected to two-way coupling analysis, and finally, a drive mechanism surface heat flux and temperature field calculation result is obtained.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation
• G06F 113/08 - Fluids
• G06F 119/08 - Thermal analysis or thermal optimisation

Abstract

The present invention relates to a detector background noise determination method and system based on fitting extrapolation. A current signal during lifting of control rods can be used to perform fitting and extrapolation data processing, so as to obtain a background noise current value suitable for dynamic rod worth measurement of each group of control rods, thereby improving the precision of a dynamic rod worth measurement test. Due to the use of a fitting extrapolation method, it is not necessary for a unit to perform an additional specific operation, and it is also not necessary to perform measurement every time a control rod is inserted, so that a key path time of a physics startup test is not increased, and an accurate background noise value during dynamic rod worth measurement of each group of control rods can be obtained.

IPC Classes  ?

• G21C 17/104 - Measuring reactivity
• G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
• G01R 31/00 - Arrangements for testing electric propertiesArrangements for locating electric faultsArrangements for electrical testing characterised by what is being tested not provided for elsewhere

Abstract

The present invention relates to a system and method for vacuuming a containment of a nuclear power plant. The system comprises a pressure vessel, wherein a containment is sleeved outside of the pressure vessel, a cavity is formed between the containment and the pressure vessel, the top of the containment is connected to each of a vacuum ejector and a vacuum pump by means of pipelines, the containment is vacuumized by means of the vacuum ejector and the vacuum pump, and a vacuum environment in the containment is established. The present invention can not only reduce the layout space of an apparatus and the pipelines, but can also reduce the heat loss of the apparatus and the pipelines, has the functions of monitoring reactor coolant pressure boundary leakage and draining water inside the containment before reactor startup, and meets the requirements of small modular reactors.

IPC Classes  ?

• G21D 1/04 - Pumping arrangements
• G21C 13/02 - Pressure vesselsContainment vesselsContainment in general Details

Abstract

A radial uniform-irradiation device and method for large-dimension monocrystalline silicon. The radial uniform-irradiation device for large-dimension monocrystalline silicon comprises: a bucket body, in which silicon ingots to be irradiated can be placed; a neutron shielding layer, arranged on the outer side of the bucket body and used for preventing neutrons from entering the bucket body, the circumference of the neutron shielding layer being provided with gaps, and the gaps being used for enabling neutrons to enter the bucket body, so as to form neutron flux areas; a reflection block, arranged at ends of silicon ingots, neutrons entering the reflection block through the gaps, and after being scattered, irradiating the silicon ingots from the end surfaces of the reflection block; and a rotation driving mechanism, provided in the bucket body and used for driving the silicon ingots to rotate. Arranging the neutron shielding layer and the reflection block at the external position of monocrystalline silicon changes the position and direction of neutrons entering silicon ingots; and rotating the silicon ingots in cooperation during irradiation achieves radial uniformity of irradiation doping for large-dimension monocrystalline silicon.

IPC Classes  ?

• C30B 31/20 - Doping by irradiation with electromagnetic waves or by particle radiation
• C30B 29/06 - Silicon

Abstract

Provided in the present invention are a creep fatigue state evaluation method and system for a high-temperature nuclear power device, the creep fatigue state evaluation method for a high-temperature nuclear power device comprising: acquiring an isochronous stress-strain curve of a device, and, according to the isochronous stress-strain curve, deducing creep data and plastic strain data of device materials; according to the creep data of the device materials, calculating a creep constitutive parameter of the device materials; according to the creep constitutive parameter, performing non-elastic creep analysis to obtain a stress parameter and a strain parameter; performing, according to the stress parameter, equivalent stress calculation to obtain a time history of a corrected equivalent stress response in a life period, and, in view of a minimum rupture stress curve, calculating a creep damage; according to the strain parameter, performing equivalent strain calculation to obtain a strain range of each time point in the life period, and, in view of a fatigue curve, calculating a fatigue damage; and, according to the creep damage and the fatigue damage, estimating a creep fatigue state of the device. The present invention is simple and easy to implement, and reduces the conservativeness and complexity of estimation by means of elastic analysis methods.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation
• G06F 119/02 - Reliability analysis or reliability optimisationFailure analysis, e.g. worst case scenario performance, failure mode and effects analysis [FMEA]
• G06F 119/04 - Ageing analysis or optimisation against ageing
• G06F 119/08 - Thermal analysis or thermal optimisation
• G06F 119/14 - Force analysis or force optimisation, e.g. static or dynamic forces

Abstract

A reactor body structure and a reactor system. The reactor body structure (100) comprises: a reactor pressure vessel (1), the reactor pressure vessel (1) comprising a bottom closure head (3), and the bottom closure head (3) having an arc-shaped inner wall (32); a reactor internal (2), the reactor internal (2) comprising a reactor core supporting lower plate (22); and a reactor core (21), the reactor core (21) being arranged on the reactor core supporting lower plate (22); wherein the reactor core supporting lower plate (22) is located on the lower side of the center of sphere of the arc-shaped inner wall (32), and the outer edge of the reactor core supporting lower plate (22) is provided with a flow guide corner (221); the reactor core supporting lower plate (22) is provided with through holes (222), the through holes (222) are stepped holes and each comprise a top hole (222a) in the upper side and a bottom hole (222b) in the lower side, and the diameter of the bottom hole (222b) is smaller than that of the top hole (222a).

IPC Classes  ?

• G21C 15/24 - Promoting flow of the coolant
• G21C 19/04 - Means for controlling flow of coolant over objects being handledMeans for controlling flow of coolant through channel being serviced

Abstract

A reactor internal of a nuclear reactor, a nuclear reactor, and a method for designing a reactor internal. The reactor internal (30) comprises a reactor core upper plate (1), a guide cylinder assembly (2), and a supporting column assembly (4). A plurality of flow holes (11) that may enable fluid to pass through are formed in the reactor core upper plate (1). The flow holes (11) comprise a first circular flow hole (111), a second circular flow hole (112), and a square flow hole (113). The guide cylinder assembly (2) is disposed above the square flow hole (113). The supporting column assembly (4) is mounted above the first circular flow hole (111). Thus, lateral flow between flow channels can be effectively reduced, abrasion between a control rod (3) and the guide cylinder assembly (2) is relieved, and the economical efficiency and safety of the nuclear reactor (100) are increased.

IPC Classes  ?

• G21C 19/04 - Means for controlling flow of coolant over objects being handledMeans for controlling flow of coolant through channel being serviced
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A mounting method and assistance device for a steel containment. The method comprises: assembling a plurality of in-place section sub-modules (6), and arranging a plurality of first supporting members (3) on the side faces of the in-place section sub-modules; arranging telescopic mechanisms (4) at the tops of the first supporting members; arranging second supporting members (2) on the side faces of hoisting section sub-modules, and using slings (20) to hoist the hoisting section sub-modules (5), so that the second supporting members and the first supporting members are correspondingly arranged, and are fitted to the output ends of the telescopic mechanisms; and adjusting the telescopic mechanisms so as to keep the bottoms of the hoisting section sub-modules horizontal, removing the slings of the hoisting section sub-modules, and assembling and welding. When the steel containment is hoisted and in place, the slings can be removed. The method and the assistance device can reduce the influences of loading of the slings on the on-site construction.

IPC Classes  ?

• B66C 13/08 - Auxiliary devices for controlling movements of suspended loads, or for preventing cable slack for depositing loads in desired attitudes or positions
• B66F 3/24 - Devices, e.g. jacks, adapted for uninterrupted lifting of loads fluid-pressure operated
• E04G 21/14 - Conveying or assembling building elements

Abstract

The present invention provides an automatic monitoring method and system for a technical specification of nuclear power plant operation. The automatic monitoring method comprises: acquiring nuclear power plant operation data, technical specification-related supervisory test procedure execution data, and a fuel circulation requirement parameter; on the basis of the nuclear power plant operation data, the technical specification-related supervisory test procedure execution data, and the fuel circulation requirement parameter, according to supervision requirements in technical specification entries, determining whether the nuclear power plant operation deviates from limiting conditions for operation; if yes, searching for a corresponding measure, and sending out early warning information and the corresponding measure to an operator so as to remind the operator to make an operation decision; and if not, continuing to execute automatic monitoring of the technical specification of the nuclear power plant operation. The problem in the prior art of monitoring of the technical specification of the nuclear power plant operation needing to be manually completed is solved.

IPC Classes  ?

• G06F 16/34 - BrowsingVisualisation therefor

Abstract

The present invention relates to a strainer failure PSA modeling method and system for a passive nuclear power plant. The method comprises the following steps: determining the total amount x of potential debris of a nuclear power plant according to the amount of inherent debris of the nuclear power plant and an expected amount of debris in chemical reaction products after an accident; analyzing a migration path of the debris according to the numbers and types of debris generated at different break positions in the nuclear power plant; by using a device reliability database of the nuclear power plant, obtaining a probability distribution of pit blockage that occurs under a certain accident type, and according to the relationship among a strainer failure rate r, the total amount x of potential debris of the nuclear power plant, a break size y and a break position z, obtaining the process of the debris being migrated to a strainer and accumulated; and dividing working conditions according to the magnitudes of x, y and z, determining the value of a strainer blockage failure rate under each working condition, and performing PSA modeling according to value results under different working conditions.

IPC Classes  ?

• G06F 30/18 - Network design, e.g. design based on topological or interconnect aspects of utility systems, piping, heating ventilation air conditioning [HVAC] or cabling
• G21C 17/00 - MonitoringTesting

Abstract

The present invention provides an adjustable joint of a modular steel structure for a large container, and a construction method. The adjustable joint comprises a connecting joint, which is arranged on a steel structure module and is provided with a slot; and a connecting device, of which one end is fixedly connected to the inner wall of a large container, and the other end can be embedded into the slot and move to a proper position in the axial direction of the slot so as to fixedly connect to the connecting joint at the proper position. During or after the construction stage of the large container, the connecting device of the adjustable joint of the present invention moves in the axial direction of the slot, so as to adapt to the deformation and stiffness of the large container, thus mitigating problems such as difficulties in assembling and welding large containers and effects of deformation on installation of connecting joints. When used for modular construction of nuclear power projects, the adjustable joint can mitigate problems that the deformation of large containers affects installation of connecting joints and causes difficulties in assembly and welding.

IPC Classes  ?

• E04B 1/38 - Connections for building structures in general
• E04B 1/24 - Structures comprising elongated load-supporting parts, e.g. columns, girders, skeletons the supporting parts consisting of metal

Abstract

bbbbb' of the molten pool, and according to a comparison result, carrying out iterative calculation to determine an actual upper boundary temperature of the top of the molten pool, an actual heat flux density of the top of the molten pool and an actual heat flux density of the side wall of the molten pool. The calculation method of the present application considers the actual situation of a power plant accident, and distinguishes two types of heat transfer models of film boiling and nucleate boiling, making the calculation result more comprehensive.

IPC Classes  ?

• G06F 30/28 - Design optimisation, verification or simulation using fluid dynamics, e.g. using Navier-Stokes equations or computational fluid dynamics [CFD]
• G06F 113/08 - Fluids
• G06F 119/08 - Thermal analysis or thermal optimisation

Abstract

The purpose of the present invention is to disclose a cast-in-place composite shielding shell with ultra-high performance concrete (UHPC), comprising an UHPC layer, rigid dismantling-free formworks, a RC layer, and a back anti-crack panel; the rigid dismantling-free formworks are disposed between the UHPC layer and the RC layer, and the RC layer is disposed between the UHPC layer and the back anti-crack panel; the RC layer and the UHPC layer are connected and fixed to each other by means of connectors. Compared with the prior art, the UHPC layer has higher strength and toughness, and is not prone to breakage and splashing under high-speed impact, and the back anti-crack panel prevents the presence of scabs on the back, and can better maintain the overall stress performance of the impacted site, thereby reducing damage to a main structure caused by the impact of a projectile, thus achieving the purpose of the present invention.

IPC Classes  ?

• E04C 3/20 - JoistsGirders, trusses, or truss-like structures, e.g. prefabricatedLintelsTransoms of concrete or other stone-like material, e.g. with reinforcements or tensioning members
• E04B 1/92 - Protection against other undesired influences or dangers

Abstract

The purpose of the present invention is to disclose a semi-fabricated composite shielding shell with ultra-high performance concrete (UHPC), comprising an UHPC layer, a RC layer, and a back anti-crack panel. The RC layer is arranged between the UHPC layer and the back anti-crack panel, and the RC layer and the UHPC layer are connected to each other by means of connectors. Compared with the prior art, the semi-fabricated composite shielding shell has the advantages that: the UHPC layer, instead of a steel plate, is used on the side of a building having a protection requirement, damage is generally limited to the UHPC layer, and main stress reinforcements of the RC layer cannot be affected; the sizes and contours of respective components (the UHPC layer, the RC layer, and the back anti-crack panel) of the shielding shell can be flexibly adjusted according to the shape and protection requirement of the building; the rigidity-density ratio can be flexibly adjusted; the thickness of the shielding shell is greatly reduced; and the space size of the building is reduced, and construction costs are reduced. The shielding shell is suitable for the protection requirement of the building against a projectile, thus the purpose of the present invention is achieved.

IPC Classes  ?

• E04B 1/94 - Protection against other undesired influences or dangers against fire

Abstract

A target used for heavy water reactor production of C-14 isotopes, which comprises: a target tube, which has a target material and/or an absorber material, the target material being a nitrogen-containing solid material; end plates, which are arranged at two ends of the target tube; a connection rod, which is provided with a connection component that is connected to a positioning mechanism used for positioning the target, wherein the connection rod passes through the target tube along the axial direction of the target tube, and the target tube is fixedly connected to the connection rod by means of the end plates. The present invention can be adapted to be placed at different locations such as at a guide tube, an inspection passage, or a fuel channel to undergo irradiation and produce C-14 isotopes.

IPC Classes  ?

• G21G 1/00 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes
• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors

Abstract

A melt transient reaction simulation device, comprising a water-cooled crucible (30) provided with a protective cover (2), the protective cover (2) being sleeved on an outer side of the water-cooled crucible (30); a cover plate (7) covering the top of the protective cover (2), wherein the cover plate (7) is provided with an air inlet pipe (73) and an exhaust pipe (74) in communication with the interior of the water-cooled crucible (30); a heating mechanism configured to heat materials in the water-cooled crucible (30); an inert gas supply mechanism connected to the air inlet pipe (73) and configured to convey into the water-cooled crucible (30) an inert gas with a density greater than that of air; an oxygen-content monitoring mechanism connected to the exhaust pipe (74) and configured to monitor the oxygen content of gas discharged from the exhaust pipe (74); and a shielding mechanism sleeved between the heating mechanism and the water-cooled crucible (30), wherein the shielding mechanism is slidably connected to the cover plate (7), and can axially move relative to the cover plate (7) so as to adjust an area, which is directly exposed to the heating mechanism, of the water-cooled crucible (30).

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G21C 17/112 - Measuring temperature
• G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain

Abstract

An external enhanced heat transfer system for a pressure vessel, which comprises: an external flow channel (3), which is used for fluid flow, and is formed by an outer wall of the pressure vessel (1) and an insulating layer (2) wrapped around the periphery of the pressure vessel (1); a coolant inlet (32) is formed at the bottom of the external flow channel (3), and a coolant outlet (33) is formed at the top of the external flow channel (3); an agitation apparatus (5), which is provided at the bottom of the external flow channel (3), and is used for agitating the fluid in the external flow channel (3); an ultrasonic vibration apparatus (4), which is arranged on the insulating layer (2), and is used for applying vibrations to the fluid in the external flow channel (3); and a nanofluid supply mechanism (6), which is in communication with the external flow channel (3), and is used for providing a nanofluid for the external flow channel (3). Agglomeration and settlement of nanoparticles are prevented, convective heat transfer is able to be enhanced under ultrasonic action, and detachment of bubbles near the surface of the outer wall of the pressure vessel (1) can also be accelerated. The critical heat flux in a boiling heat transfer process is improved, and the effectiveness of IVR measures is ensured.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel

Abstract

The present invention provides a complete measurement and processing system for a reactor core instrument, used for continuously measuring signals in a nuclear power plant reactor, and comprising: a reactor core instrument sleeve assembly, a signal processing device, and a signal cable assembly, wherein at least one reactor core instrument sleeve assembly is arranged in a reactor core; a sleeve electrical connector is provided at the tail of the reactor core instrument sleeve assembly; a thermocouple and a plurality of self-powered detectors are provided in the reactor core instrument sleeve assembly in the axial direction; the thermocouple and the core wires of the plurality of self-powered detectors are all connected to the sleeve electrical connector; one end of the signal cable assembly is provided with a cable electrical connector, and the other end of the signal cable assembly is provided with a branch mechanism; the branch mechanism is connected to a plurality of branch electrical connectors; the cable electrical connector is connected to the sleeve electrical connector; the branch electrical connectors are connected to the signal processing device. The present invention solves the problems that existing reactor core instruments are disperse, the number of the penetration interfaces of a pressure vessel is large, and the system interfaces and the connecting cables are complex, and reduces the number of the cables and a containment penetration interface.

IPC Classes  ?

• G21C 17/10 - Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
• G21C 17/112 - Measuring temperature
• G21C 17/108 - Measuring reactor flux
• G21C 17/116 - Passages or insulators, e.g. for electric cables

Abstract

A motor lead screw type control rod driving mechanism and a driving method thereof. Nuts are driven by using a controllable motor, rotation of the nuts is converted into linear motion of a lead screw (22) by means of a helical transmission pair (200), the lead screw (22) is connected to a control rod assembly (1) by means of a core rod (20), so that continuous motion and accurate positioning of the control rod assembly (1) can be achieved, and high reliability is achieved; in addition, an integral nut transmission lead screw (22) is used to always maintain correct engagement of the helical transmission pair (200), thereby avoiding the problem of too quick wear caused by engagement deviation of separable nuts, and prolonging the service life of the driving mechanism. The helical transmission pair (200) has good impact and swing resistance, and would not be separated due to emergent rod dropping or replacement, so that an impact problem caused by re-engagement and connection is avoided, engagement precision and transmission stability are ensured, the wear resistance and reliability of the transmission component are improved, and the service life of the transmission component is prolonged.

IPC Classes  ?

• G21C 7/14 - Mechanical drive arrangements

Abstract

A reactor fuel-loading and refueling system (100) and method. The method comprises performing fuel-loading and refueling operations on a reactor pressure vessel (8) having a cover capable of being opened downwards. The system comprises a bolt operating device (1), a lifting device (2), an attitude adjusting device (3), and a transfer device (4), wherein the bolt operating device (1), the lifting device (2), the attitude adjusting device (3), and the transfer device (4) are located underwater in a reactor core pool; the attitude adjusting device (3) is arranged at the top of the lifting device (2); the bolt operating device (1) is connected to the attitude adjusting device (3); the lifting device (2) is fixedly connected to the transfer device (4); and the transfer device (4) can drive the lifting device (2), and the attitude adjusting device (3) and the bolt operating device (1) that are mounted on the lifting device (2) to move in the horizontal direction. According to the reactor fuel-loading and refueling system (100) and method, the problems in the existing nuclear reactor refueling technology of a long path and high disassembly and assembly difficulty are solved, thereby shortening the path and reducing the disassembly and assembly difficulty.

IPC Classes  ?

• G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel

Abstract

The present invention provides an intelligent monitoring method and system for a nuclear power station steam generator. The system comprises a device thermal performance monitoring module, a device flow field digital twin module, a heat transfer tube flow-induced vibration monitoring module, a heat transfer tube wearing monitoring module, a device fatigue damage monitoring module, a device loosened component monitoring module, and a monitoring data and file management module, can be used for thermal performance monitoring, local flow field monitoring, heat transfer tube flow-induced vibration and wearing evaluation, and fatigue and loosened component monitoring and diagnosis of the steam generator, and is used for full life cycle management of the steam generator; in an installation process of the system, no new sensor is added, so that the installation is simpler and more convenient. Moreover, the system can monitor data such as a dirt coefficient and a flow field distribution condition which cannot be monitored, and predict the wearing condition of the heat transfer tube, so that the time required for maintenance of the steam generator is greatly shortened, and the economic benefit of a nuclear power station is improved.

IPC Classes  ?

• F22B 35/18 - Applications of computers to steam-boiler control

Abstract

The present invention provides a passive residual heat removal system and method for a nuclear reactor. The system comprises: a reactor system, comprising a hot end and a cold end, wherein the hot end outputs a fluid outwards, and the cold end inputs a fluid inwards; and a residual heat removal system, comprising a multi-stage heat exchanger, wherein every two adjacent stages of heat exchangers are respectively connected by means of an intermediate header. The hot end is connected to an inlet of a first-stage heat exchanger by means of an inlet pipeline; the intermediate header is provided with a side outlet, and the side outlet and an outlet of a last-stage heat exchanger are respectively connected to the cold end by means of outlet pipelines to form a multi-stage heat exchange loop; and the inlet pipeline and each outlet pipeline are respectively provided with an isolating valve. According to the present invention, different requirements for the heat removal capacity of a residual heat removal system under different accidents can be met, the design of the multi-stage heat exchanger avoids an excessively large or small heat removal amount, and adverse effects on a nuclear reactor system due to a false start of a passive residual heat removal system are reduced.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• F28F 13/00 - Arrangements for modifying heat transfer, e.g. increasing, decreasing

Abstract

The present application relates to a composite shielding yttrium-based alloy material. The main components of the yttrium-based alloy material comprise the following components in percent by mass: B: 0.05-10.0%, Cr≤6.0% or Al≤5.0%; and the remaining components of yttrium and other inevitable impurities. The grain size of the yttrium-based alloy material is 10-50 μm. Also provided are a preparation method for and a use of the composite shielding yttrium-based alloy material. The composite shielding yttrium-based alloy material of the present application can be used in a high-temperature environment of 600 to 1,000°C, has fine grains, relatively high strength and toughness and low density, and is a high-quality and efficient neutron moderation and absorption integrated alloy material.

IPC Classes  ?

• C22C 28/00 - Alloys based on a metal not provided for in groups
• C22C 1/02 - Making non-ferrous alloys by melting
• C22F 1/02 - Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working in inert or controlled atmosphere or vacuum
• C22F 1/16 - Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
• G21F 1/08 - MetalsAlloysCermets, i.e. sintered mixtures of ceramics and metals

Abstract

The present invention provides a steam generator, a method for relieving wear of heat transfer tubes of the steam generator, and a component mounting method for relieving wear of the heat transfers tube of the steam generator. The steam generator comprises: a housing; an inner sleeve which is arranged inside of the housing, a descending channel for fluid being formed by an outer side surface of the inner sleeve and an inner side surface of the housing; a heat transfer tube bundle which is arranged inside of the inner sleeve to exchange heat with the fluid; a primary separator which is arranged right above the heat transfer tube bundle and the inner sleeve, and is fixed inside of the housing to carry out gas-liquid separation on the fluid after heat exchange; and a resistance component which is arranged at an inlet of the descending channel to buffer the fluid flowing past and reduce the flow speed of the fluid. By means of optimizing and improving the secondary side structure of the steam generator, flow-induced vibration wear of the heat transfer tube bundle caused by an uneven flow field or excessively high flow velocity of the heat transfer tube bundle of the steam generator is overcome, effectively preventing the occurrence of heat transfer tube bundle ruptures without affecting heat transfer efficiency.

IPC Classes  ?

• G21D 1/00 - Details of nuclear power plant
• F22D 11/00 - Feed-water supply not provided for in other main groups
• F22B 1/02 - Methods of steam generation characterised by form of heating method by exploitation of the heat content of hot heat carriers

Abstract

6666, and the density of the burnable poison coating is 70%-97% of the theoretical density of the used material. The present invention further relates to a method for preparing the burnable poison coating, and a nuclear fuel element comprising a nuclear fuel pellet to which the coating is applied.

IPC Classes  ?

• C23C 14/06 - Coating by vacuum evaporation, by sputtering or by ion implantation of the coating forming material characterised by the coating material
• C23C 14/35 - Sputtering by application of a magnetic field, e.g. magnetron sputtering
• C23C 14/54 - Controlling or regulating the coating process
• G21C 3/02 - Fuel elements
• G21C 3/20 - Details of the construction within the casing with coating on fuel or on inside of casingDetails of the construction within the casing with non-active interlayer between casing and active material
• G21C 3/62 - Ceramic fuel

Abstract

The present invention belongs to the technical field of nuclear power station reactor cores, and particularly relates to an assembly for connecting a cable to a nuclear power station neutron temperature measurement channel. The present invention comprises a stainless steel corrugated pipe and a cable, wherein both ends of the stainless steel corrugated pipe are fixedly and hermetically connected to adapters by means of argon arc welding; the end of the left adapter away from the stainless steel corrugated pipe is fixedly and hermetically connected to a connecting component by means of argon arc welding; the end of the right adapter away from the stainless steel corrugated pipe is fixedly and hermetically connected to a connector by means of argon arc welding; the cable passes through a closed cavity formed by the connecting component, the stainless steel corrugated pipe and a shell of the connector; the connecting component is hermetically connected to a shell of a measurement channel; an inside cable is connected to a sensor core in the measurement channel; the shell of the connector is connected to a socket on a plug-in board beside a pool, and sealing is formed on a plug-socket insertion-connection end face; and a core of the inside cable is connected to a contact inside a plug. The present invention isolates the cable from an external high-humidity or soaking environment, thereby reducing the risk of the cable getting damp.

IPC Classes  ?

• H01B 7/282 - Preventing penetration of fluid into conductor or cable
• H02G 15/18 - Cable junctions protected by sleeves, e.g. for communication cable

Abstract

The present application relates to a dysprosium-rich nickel-tungsten alloy material for nuclear shielding, the composition thereof comprising components of the following mass percentages: C: 0.002-0.02%, W: 5.0-35.0%, Cr: 15.0-30.0%, Dy: 1.0-4.0%, and the remaining components are nickel and unavoidable impurities. A preparation method for the dysprosium-rich nickel-tungsten alloy material for nuclear shielding is also provided. In the present application, a high-dysprosium and high-tungsten nickel-tungsten alloy material is prepared by adding an appropriate ratio of nickel, chromium, tungsten and dysprosium, and has the advantages of high strength, good plasticity and toughness, corrosion resistance and excellent processing and formability, and can be used as an integrated material of a neutron and photon synergistic shielding functional structure.

IPC Classes  ?

• C22C 19/05 - Alloys based on nickel or cobalt based on nickel with chromium
• C22C 1/02 - Making non-ferrous alloys by melting
• C22C 27/04 - Alloys based on tungsten or molybdenum
• C22C 30/00 - Alloys containing less than 50% by weight of each constituent
• C22F 1/10 - Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of nickel or cobalt or alloys based thereon
• C22F 1/18 - High-melting or refractory metals or alloys based thereon
• B21C 37/02 - Manufacture of metal sheets, rods, wire, tubes, profiles or like semi-manufactured products, not otherwise provided forManufacture of tubes of special shape of sheets
• B21C 37/04 - Manufacture of metal sheets, rods, wire, tubes, profiles or like semi-manufactured products, not otherwise provided forManufacture of tubes of special shape of rods or wire
• G21F 1/08 - MetalsAlloysCermets, i.e. sintered mixtures of ceramics and metals

Abstract

The present invention provides a nuclear fuel charging/discharging and position automatic tracking system and method. The system comprises: fuel assemblies, provided with first identification information capable of uniquely identifying the fuel assemblies; inserts, arranged in the fuel assemblies and provided with second identification information capable of uniquely identifying the inserts; a system database, storing reactor core arrangement information and spent fuel pool arrangement information; and a processing apparatus, which, according to the current reactor core arrangement information and the spent fuel pool arrangement information, establishes a corresponding relationship with preset reactor core arrangement information of a next cycle, and configures shuffling movement information of each fuel assembly and a corresponding insert thereof, so as to generate a shuffling scheme. The present invention automatically generates a shuffling scheme according to the reactor core arrangement information and the spent fuel pool arrangement information in the system database of its own, so as to simplify the data processing and conversion operation during charging/discharging processes, thus improving the working efficiency, and reducing the risk of reactor core charging/discharging operation errors, and ensuring the operation safety of reactor cores.

IPC Classes  ?

• G06Q 10/04 - Forecasting or optimisation specially adapted for administrative or management purposes, e.g. linear programming or "cutting stock problem"
• G21C 19/19 - Reactor parts specifically adapted to facilitate handling, e.g. to facilitate charging or discharging of fuel elements

Abstract

An integrated reactor (100), and an integrated reactor charging and refueling system (1000) and method. The integrated reactor (100) comprises: a reactor cavity (10); a containment (1), which is arranged in the reactor cavity (10), wherein the containment (1) comprises an upper containment (11) and a lower containment (12), and the upper containment (11) and the lower containment (12) are detachably and fixedly connected; and a pressure vessel (2), which is arranged in the containment (1), wherein the pressure vessel (2) comprises an upper pressure vessel (21) and a lower pressure vessel (22), the upper pressure vessel (21) and the lower pressure vessel (22) are detachably and fixedly connected, and the upper pressure vessel (21) and the upper containment (11) are fixedly connected to form an integrated hoisting structure (20).

IPC Classes  ?

• G21C 19/20 - Arrangements for introducing objects into the pressure vesselArrangements for handling objects within the pressure vesselArrangements for removing objects from the pressure vessel
• G21C 19/07 - Storage racksStorage pools
• G21C 19/10 - Lifting devices or pulling devices adapted for co-operation with fuel elements or with control elements

Abstract

Integrated passive reactor including a reactor primary circuit, a containment cooling system, a residual heat removal system, and a reactor core cooling system. Loop resistance is reduced by means of a reactor-type process design, a flow guide device is provided at a rising section of fluid to reduce the loop resistance, the rising section is shrunken to increase the arrangement space of a heat exchanger so as to further optimize system resistance, and the designs of an infinite-time passive reactor core residual heat removal system and an infinite-time passive containment cooling system are achieved. By means of the rational configuration of a pressure relief system, high-pressure safety injection is removed, and the passive reactor core cooling system is simplified. By means of the design of an auxiliary circulation device for a loss of coolant accident, the safety of a reactor core in the loss of coolant accident is further enhanced.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel

Abstract

An alkali metal reactor power supply, including: a reactor vessel, the bottom part of which is provided with a liquid alkali metal; a reactor core, which is arranged in the reactor vessel and includes a plurality of fuel rods and a radial reflection layer arranged at the periphery of the plurality of fuel rods, wherein the surface of each fuel rod is provided with a first liquid absorption core, the bottom part of the reactor core is provided with second liquid absorption cores which are connected to the first liquid absorption cores, and the second liquid absorption cores can be in contact with the liquid alkali metal; and alkali metal thermoelectric converters, which are arranged along the circumferential direction of the radial reflection layer, and divide the inside of the reactor vessel into a high-pressure steam chamber located above the alkali metal thermoelectric converters and a low-pressure steam chamber located below the alkali metal thermoelectric converters. By using the phase-change heat transfer of alkali metal, the circulating power of the liquid alkali metal is provided by using the liquid absorption cores, the structure is simple, the arrangement is flexible, and the power generation efficiency is high.

IPC Classes  ?

• G21D 7/04 - Arrangements for direct production of electric energy from fusion or fission reactions using thermoelectric elements
• G21C 3/04 - Constructional details
• H10N 10/854 - Thermoelectric active materials comprising inorganic compositions comprising only metals

Abstract

A system for producing radioactive isotopes by using a heavy water reactor nuclear power plant, comprising: a production channel (2) comprising a straight pipe section (211), a bent pipe section (212) and an inclined pipe section (213) which are connected together in sequence, wherein the bottom of the straight pipe section (211) is located at the bottom of the inner side of a heavy water reactor calandria vessel (8), the inclined pipe section (213) is provided with a charging port (21) and a discharging port (22), and the charging port (21) is closer to the bent pipe section (212) than the discharging port (22); a target box carrier (3) provided in the production channel (2) and used for bearing a target box (4) and driving the target box (4) to move in the production channel (2); a target box traction mechanism (1) provided at the tail end of the inclined pipe section (213) away from the bent pipe section (212) and used for pulling the target box carrier (3); an automatic transport mechanism provided above the calandria vessel (8) and used for transporting the target box (4) before irradiation production and receiving the irradiated target box (4) in the production channel (2); and a conveying mechanism (10) used for conveying the target box (4) before irradiation production to the charging port (21) and placing the target box (4) in the target box carrier (3). Mutual impact of the target box (4) is avoided, radiation borne by the target box traction mechanism (1) is reduced, and disturbance to a reactor core is reduced.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors

Abstract

Provided in the present application is a reactor power supply system. The reactor power supply system comprises an alternating current source system, which comprises a conventional alternating current source; and a direct current source system, wherein the conventional alternating current source comprises a high-voltage bus, a low-voltage bus and a voltage transformation device, the high-voltage bus is powered by a small reactor module, and the high-voltage bus supplies power to the low-voltage bus after voltage transformation is performed by the voltage transformation device; and the direct current source system comprises an alternating-direct current converter, a direct-current bus and a storage battery, and the alternating-direct current converter is connected to the low-voltage bus, so as to convert an alternating current that is transmitted by the low-voltage bus into a direct current to charge the storage battery by means of the direct-current bus. In the present invention, the comprehensive utilization of nuclear energy of a small reactor and a rational power supply system design are taken into comprehensive consideration, thereby expanding the application scenarios of the small reactor, and improving the reliability of nuclear energy power supply. A passive design of an advanced small reactor is matched, and a safety-class power source device does not have to be used, thereby further improving the safety and economical efficiency of the small reactor.

IPC Classes  ?

• H02J 4/00 - Circuit arrangements for mains or distribution networks not specified as ac or dc

Abstract

A nuclear island base slab of a nuclear power plant and a manufacturing method therefor, and a nuclear island of a nuclear power plant. The nuclear island base slab of a nuclear power plant includes a concrete base slab body and a plurality of air ducts embedded in the concrete base slab body. The air duct has an internal-penetrating bent pipe structure. A first end of the air duct is exposed on an upper surface of the concrete base slab body. A second end of the air duct is exposed on a side surface or the upper surface of the concrete base slab body.

IPC Classes  ?

• E02D 27/00 - Foundations as substructures
• E02D 15/02 - Handling of bulk concrete specially for foundation purposes
• E04G 21/02 - Conveying or working-up concrete or similar masses able to be heaped or cast
• E21F 1/04 - Air ducts

Abstract

Provided is a reactor secondary side passive residual heat removal system, comprising: a containment vessel; a steam generator provided with a steam outlet and a water supply inlet; a water tank, the water tank being internally provided with a heat exchanger, the heat exchanger having a heat exchanger inlet and a heat exchanger outlet; and a steam driven pump provided with a steam port, a water inlet and a water outlet, wherein the steam generator, the water tank and the steam driven pump are arranged in the containment vessel, the heat exchanger inlet is in communication with the steam outlet of the steam generator by means of a first pipeline, the heat exchanger outlet is in communication with the water inlet of the steam driven pump by means of a second pipeline, the water outlet of the steam driven pump is in communication with the water supply inlet of the steam generator by means of a third pipeline, and the steam port of the steam driven pump is in communication with the first pipeline by means of a fourth pipeline. The present invention does not rely on an external driving force, thereby greatly reducing the failure probability of the system and improving the safety of the system.

IPC Classes  ?

• G21C 15/12 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from pressure vesselArrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements from containment vessel
• G21C 15/243 - Promoting flow of the coolant for liquids
• G21C 17/032 - Reactor-coolant flow measuring or monitoring

Abstract

A passive waste heat removal system on the secondary side of a marine environmental reactor. The system includes a containment, an airtight water tank, and a steam generator. The containment is partially or fully immersed in seawater. The airtight water tank is disposed on the inner wall surface of the containment, the airtight water tank being provided with a water tank inlet and a water tank outlet. The steam generator is placed in the containment, the steam generator having a steam outlet and a feedwater inlet. The water tank inlet of the airtight water tank communicates with the steam outlet of the steam generator by means of a first pipe, and the water tank outlet of the airtight water tank communicates with the feedwater inlet of the steam generator by means of a second pipe.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A method for designing components of a nuclear power-use nickel-based alloy welding wire for preventing welding cracks, a nuclear power-use nickel-based alloy welding wire, and a method for detecting welding cracks. Crystallization cracks are prevented by controlling the volume percentage of the microstructure Laves phase.

IPC Classes  ?

• B23K 35/30 - Selection of soldering or welding materials proper with the principal constituent melting at less than 1550°C
• B23K 35/40 - Making wire or rods for soldering or welding
• C22C 19/05 - Alloys based on nickel or cobalt based on nickel with chromium

Abstract

The present invention provides a composite seismic isolation and absorption system for a nuclear island structure, comprising: a box body having a box body bottom plate and a box body side wall, the nuclear island structure being placed in the box body, and the nuclear island structure having a nuclear island structure bottom plate and a nuclear island structure side wall; a plurality of seismic isolation supports placed between the box body bottom plate and the nuclear island structure bottom plate; and a plurality of seismic absorption devices placed between the box body side wall and the nuclear island structure side wall. The composite seismic isolation and absorption system for a nuclear island structure of the present invention greatly improves the safety of the nuclear island structure under the action of earthquakes and the ability to resist earthquakes.

IPC Classes  ?

• E04B 1/98 - Protection against other undesired influences or dangers against vibrations or shocksProtection against other undesired influences or dangers against mechanical destruction, e.g. by air-raids
• E04H 9/02 - Buildings, groups of buildings or shelters adapted to withstand or provide protection against abnormal external influences, e.g. war-like action, earthquake or extreme climate withstanding earthquake or sinking of ground
• E02D 27/32 - Foundations for special purposes
• G21C 13/024 - Supporting constructions for pressure vessels or containment vessels

Abstract

The present invention relates to the technical field of pipe vibration isolation devices, and relates in particular to a fully-coated two-stage shock absorber, which comprises a first-stage vulcanized pipe clamp assembly which fixes an upper pipe clamp and a lower pipe clamp by means of a pipe clamp connecting fastener so as to coat an upper lining rubber layer and a lower lining rubber layer therein; and further comprises rubber blocks for vibration isolation which are symmetrically arranged on the left and right sides of a bottom surface of the lower pipe clamp; and a second-stage low-frequency vibration isolation assembly which is composed of a V-shaped support seat and which is fixed to the bottom surfaces of the rubber blocks for vibration isolation, wherein the pipe clamp connecting fastener and the mounting position thereof are each coated with epoxy resin anti-corrosion coatings, and the remainder are subject to vulcanization treatment. Compared with the exiting technology, a fully-coated structure and a two-stage vibration isolation structure are used in the present invention, so that a better vibration isolation effect may be achieved, the amount of space occupied is reduced, and the weight of the components is reduced, and the anti-corrosion and insulation effect is also improved, and the service life of components is prolonged; in addition, the present invention has good application prospects in a ship system, and may also be popularized to nuclear power, petroleum and chemical engineering and other industries.

IPC Classes  ?

• F16L 3/10 - Supports for pipes, cables or protective tubing, e.g. hangers, holders, clamps, cleats, clips, brackets substantially surrounding the pipe, cable or protective tubing divided, i.e. with two members engaging the pipe, cable or protective tubing
• F16L 55/02 - Energy absorbersNoise absorbers
• F16L 55/035 - Noise absorbers in the form of specially adapted hangers or supports

Abstract

The present invention relates to the technical field of automatic depressurizing systems for steam in nuclear power plants, and in particular, to a novel automatic depressurizing system and method for a nuclear power plant, which is a passive design. The novel automatic depressurizing system comprises a voltage stabilizer, an ADS isolation valve, an ejector, an upper water tank located in a containment, a bubbler, a lower water tank, an ADS discharge pipeline, and a water suction pipeline. An outlet end of the voltage stabilizer is sequentially connected to the ADS isolation valve, one water inlet of the ejector, and the bubbler by using the ADS discharge pipeline. The bubbler is located in the upper water tank, and the lower water tank is connected to the other water inlet of the ejector by using the water suction pipeline. Compared with the prior art, in the present invention, energy released during steam ejection can be effectively utilized, such that the energy originally discharged directly into the upper water tank is effectively utilized, and meanwhile, a water source which is not used in the lower water tank originally is pumped into the upper water tank, and the size of the upper water tank can also be reduced, such that the arrangement space inside the containment is utilized to the maximum extent, and the safety and economy of the nuclear power plant are finally improved.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 9/004 - Pressure suppression
• G21C 15/25 - Promoting flow of the coolant for liquids using jet pumps

Abstract

An integrated passive reactor system including a pressure vessel, a containment vessel arranged outside the pressure vessel, and a reactor core arranged inside the pressure vessel. A primary loop operates in full natural circulation. The reactor system is provided with a secondary side passive residual heat removal system including a primary loop heat exchanger arranged inside the pressure vessel and a passive residual heat removal heat exchanger arranged outside the containment vessel. The primary loop heat exchanger is arranged above the reactor core. The passive residual heat removal heat exchanger is arranged inside a water tank which is fixed outside the containment vessel. The primary loop heat exchanger and the passive residual heat removal heat exchanger are connected by heat exchanger inlet pipelines and heat exchanger outlet pipelines.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat
• G21C 1/32 - Integral reactors, i.e. reactors wherein parts functionally associated with the reactor but not essential to the reaction, e.g. heat exchangers, are disposed inside the enclosure with the core
• G21C 13/024 - Supporting constructions for pressure vessels or containment vessels
• G21C 1/09 - Pressure regulating arrangements, i.e. pressurisers

Abstract

A target used for heavy water reactor production of C-14 isotopes, which comprises: a target tube, which has a target material (11) and/or an absorber material (14), the target material (11) being a nitrogen-containing solid material; end plates (2), which are arranged at two ends of the target tube; a connection rod (3), which is provided with a connection component that is connected to a positioning mechanism used for positioning the target, wherein the connection rod (3) passes through the target tube along the axial direction of the target tube, and the target tube is fixedly connected to the connection rod (3) by means of the end plates (2). The present invention can be adapted to be placed at different locations such as at a guide tube, an inspection passage, or a fuel channel to undergo irradiation and produce C-14 isotopes.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors

Abstract

A target used for heavy water reactor production of C-14 isotopes, which comprises: a target tube, which has a target material (11) and/or an absorber material (14), the target material (11) being a nitrogen-containing solid material; end plates (2), which are arranged at two ends of the target tube; a connection rod (3), which is provided with a connection component that is connected to a positioning mechanism used for positioning the target, wherein the connection rod (3) passes through the target tube along the axial direction of the target tube, and the target tube is fixedly connected to the connection rod (3) by means of the end plates (2). The present invention can be adapted to be placed at different locations such as at a guide tube, an inspection passage, or a fuel channel to undergo irradiation and produce C-14 isotopes.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors

Abstract

A pressure relief valve system and a pressure relief method. The system comprises a main hydraulic valve (1) and a trigger unit (4), wherein the opening and closing of the main hydraulic valve (1) is controlled by the trigger unit (4), such that a liquid in a high-pressure container (2) flows into a low-pressure container (3). By means of the pressure relief valve system and the pressure relief method, a power supply is not needed for driving, the pressure relief valve system is in an initial closed state by means of hydraulic pressure, and is passively opened after signal triggering. Therefore, the design is simplified, the safety and economical efficiency of a reactor are greatly improved, and quick pressure relief of a high-pressure container can be realized, so as to meet a long-term emergency reactor core cooling function.

IPC Classes  ?

• F16K 17/10 - Safety valvesEqualising valves opening on surplus pressure on one sideSafety valvesEqualising valves closing on insufficient pressure on one side spring-loaded with auxiliary valve for fluid operation of the main valve
• F16K 31/06 - Operating meansReleasing devices electricOperating meansReleasing devices magnetic using a magnet

Abstract

The present application relates to an integrated passive reactor, comprising a reactor primary circuit, a containment cooling system, a residual heat removal system, and a reactor core cooling system. According to the present application, loop resistance is reduced by means of a reactor-type process design, a flow guide device is provided at a rising section of fluid to reduce the loop resistance, the rising section is shrunken to increase the arrangement space of a heat exchanger so as to further optimize system resistance, and the designs of an infinite-time passive reactor core residual heat removal system and an infinite-time passive containment cooling system are achieved. By means of the rational configuration of a pressure relief system, high-pressure safety injection is removed, and the passive reactor core cooling system is simplified. By means of the design of an auxiliary circulation device for a loss of coolant accident, the safety of a reactor core in the loss of coolant accident is further enhanced. According to the integrated passive reactor provided by the present application, safety system configuration is simplified, a safety-grade alternating-current power supply is omitted, infinite-time cooling of the reactor and the containment is achieved, intervention of an operator is not required during an accident, and the safety and economy of a power plant are improved.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

The present invention provides an irradiation target for producing a Mo-99 isotope in a heavy water reactor, comprising: a plurality of fuel elements, each comprising a cladding, a uranium-containing core, and an end plug. An end part of the cladding is provided with a filling opening. The end plug covers the filling opening at the end part of the cladding, and is in sealing connection with the cladding. The uranium-containing core is filled in a sealed cavity formed by the end plug and the cladding. A plurality of fuel elements are provided. End plates are arranged at both ends of the fuel elements and are fixedly connected to the plurality of fuel elements. The uranium-containing core of at least one of the fuel elements is an uranium-containing core provided with a rich uranium fuel, and the 235 U enrichment degree of enriched uranium fuel is 6.0 wt% - 20.0 wt%. By means of the present invention, existing reactors may be utilized for non-stop production of Mo-99 having a short half-life. By using enriched uranium to produce Mo-99, the efficiency is high and the quality is good. Meanwhile, effects on power generation of a nuclear power plant can be reduced to a maximum extent.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors
• G21K 5/08 - Holders for targets or for objects to be irradiated

Abstract

The present invention provides an irradiation target for producing a Mo-99 isotope in a heavy water reactor, comprising: a plurality of fuel elements, each comprising a cladding, a uranium-containing core, and an end plug. An end part of the cladding is provided with a filling opening. The end plug covers the filling opening at the end part of the cladding, and is in sealing connection with the cladding. The uranium-containing core is filled in a sealed cavity formed by the end plug and the cladding. A plurality of fuel elements are provided. End plates are arranged at both ends of the fuel elements and are fixedly connected to the plurality of fuel elements. The uranium-containing core of at least one of the fuel elements is an uranium-containing core provided with a rich uranium fuel, and the 235 U enrichment degree of enriched uranium fuel is 6.0 wt% - 20.0 wt%. By means of the present invention, existing reactors may be utilized for non-stop production of Mo-99 having a short half-life. By using enriched uranium to produce Mo-99, the efficiency is high and the quality is good. Meanwhile, effects on power generation of a nuclear power plant can be reduced to a maximum extent.

IPC Classes  ?

• G21G 1/02 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes in nuclear reactors
• G21G 1/00 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes
• G21K 5/08 - Holders for targets or for objects to be irradiated

Abstract

An irradiation target containing a support rod for producing a molybdenum-99 isotope in a heavy water reactor, comprising: a plurality of fuel elements (1), at least one of which comprising a support rod (14) having at least two through holes, the through holes being arranged along an axial direction of the support rod (14); enriched uranium cores (13) provided in the through holes; end plates (2) provided at both ends of the fuel elements (1) and fixedly connected to the plurality of fuel elements (1). A plurality of fuel elements (1) are provided. The enriched uranium cores (13) use a rich uranium fuel having a 235U enrichment degree 15.0 wt% - 20.0 wt%. By means of the present invention, existing reactors may be utilized for non-stop production of 99Mo having a short half-life. By using enriched uranium to produce 99Mo, the efficiency is high and the quality is good. Meanwhile, effects on power generation of a nuclear power plant can be reduced to a maximum extent.

IPC Classes  ?

• G21G 1/08 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes outside of nuclear reactors or particle accelerators by neutron irradiation accompanied by nuclear fission
• G21K 5/08 - Holders for targets or for objects to be irradiated

Abstract

An irradiation target containing a support rod for producing a molybdenum-99 isotope in a heavy water reactor, comprising: a plurality of fuel elements (1), at least one of which comprising a support rod (14) having at least two through holes, the through holes being arranged along an axial direction of the support rod (14); enriched uranium cores (13) provided in the through holes; end plates (2) provided at both ends of the fuel elements (1) and fixedly connected to the plurality of fuel elements (1). A plurality of fuel elements (1) are provided. The enriched uranium cores (13) use a rich uranium fuel having a 235U enrichment degree 15.0 wt% - 20.0 wt%. By means of the present invention, existing reactors may be utilized for non-stop production of 99Mo having a short half-life. By using enriched uranium to produce 99Mo, the efficiency is high and the quality is good. Meanwhile, effects on power generation of a nuclear power plant can be reduced to a maximum extent.

IPC Classes  ?

• G21G 1/08 - Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation, or particle bombardment, e.g. producing radioactive isotopes outside of nuclear reactors or particle accelerators by neutron irradiation accompanied by nuclear fission
• G21C 3/00 - Reactor fuel elements or their assembliesSelection of substances for use as reactor fuel elements
• G21K 5/08 - Holders for targets or for objects to be irradiated

Abstract

A passive safety system for a reactor, comprising a pressure vessel, a reactor compartment (1) disposed outside the pressure vessel, and a reactor core and a main loop heat exchanger disposed inside the pressure vessel. The pressure vessel comprises a first-layer housing (13) and a second-layer housing (5) disposed at the top of the first-layer housing (13). A double-layer structure is formed at the top of the first-layer housing (13). The passive safety system for a reactor is further provided with a passive waste heat exchanger (10) and a water supply tank (15). The main loop heat exchanger (9) is disposed above the reactor core. The main loop heat exchanger (9) and the passive waste heat exchanger (10) are connected to each other by means of a heat exchange inlet pipe (8) and a heat exchange outlet pipe (11) to form a heat exchange loop. One end of the water supply tank (15) is connected to the heat exchange inlet pipe (8), and the other end is connected to the heat exchange outlet pipe (11), so that the water supply tank (15) and the passive waste heat exchanger (10) form a parallel loop. The passive safety system for a reactor makes full use of the infinite operating environment of the ocean or atmosphere to effectively deal with design basis accidents of nuclear power plants, thereby ensuring the safety of the reactor; moreover, equipment is simplified to the greatest extent, thereby improving the economy.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A system and method for verifying the control logic design of a nuclear power plant. The system comprises a client computer (10), a control logic design verification data storage server (11) and a data structure decomposition server (12), wherein the client computer (10) is respectively in communication connection with the control logic design verification data storage server (11) and the data structure decomposition server (12) by means of a network (14); the client computer (10) is used for installing a control logic design verification main program, and sending and executing all requests and actions during a control logic design verification process; the control logic design verification data storage server (11) comprises control logic design verification data; the data structure decomposition server (12) is used for loading the control logic design verification data from the control logic design verification data storage server (11), and parsing the control logic design verification data; and the control logic design verification main program performs data verification on the parsed control logic design verification data. By means of the method, automation of control logic design verification can be realized, thereby reducing the cost of labor and the human error rate.

IPC Classes  ?

• G06F 30/20 - Design optimisation, verification or simulation

Abstract

An alkali metal reactor power supply, comprising: a reactor vessel (1), the bottom part of which is provided with a liquid alkali metal (2); a reactor core (4), which is arranged in the reactor vessel (1) and comprises a plurality of fuel rods (3) and a radial reflection layer (11) arranged at the periphery of the plurality of fuel rods (3), wherein the surface of each fuel rod (3) is provided with a first liquid absorption core (5), the bottom part of the reactor core (4) is provided with second liquid absorption cores (19) which are connected to the first liquid absorption cores (5), and the second liquid absorption cores (19) can be in contact with the liquid alkali metal (2); and alkali metal thermoelectric converters (7), which are arranged along the circumferential direction of the radial reflection layer (11), and divide the inside of the reactor vessel (1) into a high-pressure steam chamber (6) located above the alkali metal thermoelectric converters (7) and a low-pressure steam chamber (8) located below the alkali metal thermoelectric converters (7). By using the phase-change heat transfer of alkali metal, the circulating power of the liquid alkali metal (2) is provided by using the liquid absorption cores, the structure is simple, the arrangement is flexible, and the power generation efficiency is high.

IPC Classes  ?

• G21C 15/02 - Arrangement or disposition of passages in which heat is transferred to the coolant, e.g. for coolant circulation through the supports of the fuel elements
• H02N 3/00 - Generators in which thermal or kinetic energy is converted into electrical energy by ionisation of a fluid and removal of the charge therefrom

Abstract

A moisture separator unit (100) of a moisture separator reheater, the moisture separator unit comprising: a drain tank (1) and two separation components (2), wherein the two separation components (2) are symmetrically arranged on the drain tank (1), such that the moisture separator unit (100) is in the shape of the letter "Y". Each separation component (2) is composed of a frame component (21) and a corrugated plate component (22), wherein the corrugated plate component (22) is fixed in the frame component (21); the frame component (21) is provided with a steam inlet (215) and a steam outlet (216), the steam inlet (215) being provided with a gas equalizing hole plate (23); and the bottom of the frame component (21) is provided with a through hole (2121), the through hole (2121) being in communication with the drain tank (1).

IPC Classes  ?

• B01D 53/26 - Drying gases or vapours
• B01D 45/08 - Separating dispersed particles from gases or vapours by gravity, inertia, or centrifugal forces by utilising inertia by impingement against baffle separators
• B01D 45/12 - Separating dispersed particles from gases or vapours by gravity, inertia, or centrifugal forces by centrifugal forces

Abstract

The present application relates to a nuclear power plant nuclear island base slab, a manufacturing method therefor, and a nuclear power plant nuclear island. The nuclear power plant base slab comprises a concrete slab body and multiple gas guiding pipes, buried within the concrete slab body, where the gas guiding pipes are internally-through bent pipe structures, the first end parts of the gas guiding pipes are exposed on the upper surface of the concrete slab body, and the second end parts of same are exposed on a lateral surface or the upper surface of the concrete slab body. With the gas guiding pipes being buried in the concrete slab body, when the concrete slab body is being casted and maintained, the heat in the nuclear island concrete slab body is removed by utilizing the natural convection in the gas guiding pipes, the rate of cooling inside the concrete is increased, and the temperature difference between the outer surface and the interior of the concrete is reduced, thus reducing the time spent on concrete maintenance, reducing the possibility of severe cracks being produced on the concrete surface, and when concrete maintenance is completed, a grouting technique is used to grout and compact the gas guiding pipes, and because of the presence of a sinking height, grouting and compaction are facilitated.

IPC Classes  ?

• E02D 15/02 - Handling of bulk concrete specially for foundation purposes

Abstract

The present invention provides a bottom closure of a reactor pressure vessel and a preparation method therefor. The bottom closure of a reactor pressure vessel comprises a bottom closure body and a porous coating applied on the outer wall surface of the bottom closure body. An end portion of the bottom closure body is connected to a barrel section of a reactor pressure vessel. According to the bottom closure of the reactor pressure vessel and the preparation method therefor in the present invention, the critical heat flux density of the outer wall surface of the reactor pressure vessel is increased, thereby facilitating the implementation of an in-vessel molten material retention technology and improving the safety of nuclear power plants even in severe accident scenarios.

IPC Classes  ?

• G21C 13/028 - Seals, e.g. for pressure vessels or containment vessels
• G21C 13/06 - Sealing-plugs

Abstract

An acousto-optic leakage monitoring system for main steam pipeline in nuclear power plant. The system includes an acoustic emission leakage monitoring loop and a spectrum leakage monitoring loop, wherein the signal input ends of the acoustic emission leakage monitoring loop and the signal input ends of the spectrum leakage monitoring loop are respectively arranged at detection points of the main steam pipeline. The signal output ends of the acoustic emission leakage monitoring loop and the signal output ends of the spectrum leakage monitoring loop are communicatively connected to each other through a network switch, and the network switch is sequentially connected with a control unit and a display unit. Compared with the prior art, the acousto-optic leakage monitoring system for the main steam pipeline in nuclear power plant according to the present invention provides early warning before the main steam pipeline leaks and realizes the diversity and redundancy of the main steam pipeline leakage monitoring methods by combining acoustic emission and spectroscopy, so that the detection results are more credible, and the maintenance cost after installation is extremely low. The detection sensitivity is higher and the response time is shortened, which significantly improves the response speed after leakage is found and provides a larger safety margin.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations
• F17D 5/00 - Protection or supervision of installations
• F17D 5/06 - Preventing, monitoring, or locating loss using electric or acoustic means

Abstract

Provided is a reactor secondary side passive residual heat removal system, comprising: a containment vessel; a steam generator provided with a steam outlet and a water supply inlet; a water tank, the water tank being internally provided with a heat exchanger, the heat exchanger having a heat exchanger inlet and a heat exchanger outlet; and a steam driven pump provided with a steam port, a water inlet and a water outlet, wherein the steam generator, the water tank and the steam driven pump are arranged in the containment vessel, the heat exchanger inlet is in communication with the steam outlet of the steam generator by means of a first pipeline, the heat exchanger outlet is in communication with the water inlet of the steam driven pump by means of a second pipeline, the water outlet of the steam driven pump is in communication with the water supply inlet of the steam generator by means of a third pipeline, and the steam port of the steam driven pump is in communication with the first pipeline by means of a fourth pipeline. The present invention does not rely on an external driving force, thereby greatly reducing the failure probability of the system and improving the safety of the system.

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A passive waste heat removal system on the secondary side of a marine environmental reactor, comprising: a containment (1) that is partially or fully immersed in seawater; an airtight water tank (4) that is disposed on the inner wall surface of the containment (1), the airtight water tank (4) being provided with a water tank inlet (41) and a water tank outlet (42); and a steam generator (7) that is placed in the containment (1), the steam generator (7) having a steam outlet (71) and a feedwater inlet (72), wherein the water tank inlet (41) of the airtight water tank (4) communicates with the steam outlet (71) of the steam generator (7) by means of a first pipe (2), and the water tank outlet (42) of the airtight water tank (4) communicates with the feedwater inlet (72) of the steam generator (7) by means of a second pipe (5).

IPC Classes  ?

• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

A nuclear fuel transport container (100), comprising an inner housing (1), a first accommodating space (11) for accommodating a nuclear fuel assembly being provided in said housing; an outer housing (2), a second accommodating space (21) for accommodating at least one inner housing (1) being provided within said outer housing; an inner/outer housing linking apparatus (3), which connects the inner housing (1) to the outer housing (2); wherein a neutron absorption plate (12) is provided at a side of the inner housing (1) facing the nuclear fuel assembly, the outer housing (2) is a two layer housing body structure made up of an outer layer housing body (22) facing an outside atmosphere side and an inner layer housing body (23) facing an inner housing (1) side, an enclosed space is formed by the outer layer housing body (22) and the inner layer housing body (23), a buffering thermal insulation material (24) is filled into the enclosed space, and a neutron moderation plate (4) is provided between the inner layer housing body (23) and the inner housing (1). The present transport container (100) ensures the safety of a nuclear fuel assembly when subject to shaking impact, allowing the nuclear fuel assembly to remain in a sub-critical state; also, a single outer housing (2) may accommodate a plurality of inner housings (1), and transportation of a plurality of nuclear fuel assemblies can be accomplished with only a single round of assembly and disassembly of the outer housing (2), improving transportation efficiency.

IPC Classes  ?

• G21F 5/08 - Shock-absorbers, e.g. impact buffers for containers

Abstract

An antishock cabinet, comprising: frame beams (1) each provided with an opening; first reinforcing members (3) embedded in the openings of adjacent frame beams (1) and connecting adjacent frame beams (1) perpendicularly to each other; columns (2) screwed or riveted to the bottom walls of the openings of the frame beams (1); and second reinforcing members (4) embedded in the cavities of the columns (2). The frame beams (1) of the cabinet have strong resistance to deformation; the columns (2) are not easy to twist; the entire cabinet structure has high antishock strength; and the second reinforcing members (4) do not occupy the internal volume of the cabinet, so that the utilization rate of the internal volume of the cabinet is high.

IPC Classes  ?

• H02B 1/01 - Frameworks
• H02B 1/54 - Anti-seismic devices or installations

Abstract

Provided is an integrated passive reactor system (100), comprising a pressure vessel (1), a containment vessel (2) arranged outside the pressure vessel (1), and a reactor core (3) arranged within the pressure vessel (1), a main circuit runs in full natural circulation. The integrated passive reactor system (100) is further provided with a secondary side passive waste heat discharge system, the secondary side passive waste heat discharge system comprises a main heat exchanger (41) arranged in the pressure vessel (1), a passive waste heat dissipation heat exchanger (42) arranged outside the containment vessel (2), and the main heat exchanger (41) is arranged above the core (3), the passive waste heat dissipation heat exchanger (42) is located inside a water tank (43) fixed outside the containment vessel (2), the main heat exchanger (41) and the passive waste heat dissipation heat exchanger (42) are connected by an inlet line (44) and an outlet line (45) of the heat exchanger. The passive safety technology, passive waste heat removal system, a double-layer structure on the top of the pressure vessel (1) and break isolation measures are used to minimize the loss of a coolant, so that it can meet the design basis accident mitigation requirements and ensure the safety of the reactor, as well as simplify the system design.

IPC Classes  ?

• G21C 13/024 - Supporting constructions for pressure vessels or containment vessels
• G21C 13/028 - Seals, e.g. for pressure vessels or containment vessels
• G21C 15/18 - Emergency cooling arrangementsRemoving shut-down heat

Abstract

Provided in the present solution is a multi-switch interlocking apparatus, related to the technical field of power supply and power distribution apparatus, comprising a first rotating rod used for controlling the opening or closing of an end switch, a second rotating rod used for controlling the opening or closing of another end switch, a first connecting rod mechanism rotatably connected to the first rotating rod and used for locking the second rotating rod, and a second connecting rod mechanism rotatably connected to the second rotating rod and used for locking the first rotating rod. The multi-switch interlocking apparatus of the present solution implements multiple switch interlocking modifications and adjustments, implements the goal of other switches being locked and prevented from being opened when one switch is opened, thus preventing a short circuit accident from be caused by an inadvertent operation of opening other switches when a switch is opened. The structure is compact and simple to install and operate; moreover, by using a mechanical interlocking scheme, the remaining switches are locked when one switch is opened, safeness and stability are improved.

IPC Classes  ?

• H01H 9/26 - Interlocking, locking, or latching mechanisms for interlocking two or more switches

Abstract

An acousto-optic leakage monitoring system for a nuclear power plant main steam pipeline (1). The system comprises an acoustic emission leakage monitoring loop (10) and an optical spectrum leakage monitoring loop (20), a signal input end of the acoustic emission leakage monitoring loop (10) and a signal input end of the optical spectrum leakage monitoring loop (20) are respectively disposed at probing points (2) of the main steam pipeline (1), a signal output end of the acoustic emission leakage monitoring loop (10) and a signal output end of the optical spectrum leakage monitoring loop (20) are in communication connection with each other by means of a network switch (30), and a control unit (40) and a display unit (50) are sequentially connected to the network switch (30). By combining acoustic emission and spectroscopy, pre-warning is made before the leakage of the main steam pipeline (1), and the diversity and redundancy of the leakage monitoring method of the main steam pipeline (1) are implemented, so that the monitoring result is more credible; and the system is lower in maintenance costs after mounting, is high in monitoring sensitivity, fast in response, significantly improve the response speed after leakage is discovered, and provide large safety allowance.

IPC Classes  ?

• G21C 17/00 - MonitoringTesting
• G21C 17/017 - Inspection or maintenance of pipe-lines or tubes in nuclear installations