938results about "Nuclear reaction control" patented technology

An integral molten salt nuclear reactor includes a drain tank section configured to hold a volume of fuel salt. The integral molten salt nuclear reactor further includes a reactor section configured to receive the volume of fuel salt from the drain tank and cause fission reactions that heats the molten salt. The integral molten salt nuclear reactor further includes a heat exchange section configured to receive a flow of the heated fuel salt from the reactor section and remove heat therefrom.

The invention belongs to the technical field of preparation of boron carbide ceramics, and particularly relates to a preparation method of high-purity porous boron carbide ceramics based on high-frequency induction heating, under the condition that a sintering aid is not added, a plunger is forced to move downwards due to volume expansion caused by CO2 generated by a melt, so that boron carbide powder is pressurized, the effect of hot pressing sintering can be achieved, and the effect of high-purity porous boron carbide ceramics can be achieved. The problems of low yield and high cost caused by the use of a hot-pressing sintering furnace can be avoided, the comprehensive performance of the prepared boron carbide ceramic can be improved, the process is easy to control, and the method is suitable for mass production of the boron carbide ceramic round rod for the high-temperature gas cooled reactor nuclear control rod; the prepared boron carbide ceramic has the advantages of pure phase, high porosity, uniform pore structure and stable mechanical property, especially has very good pore connectivity, and can ensure timely discharge of helium and improve the safety of the nuclear control rod when being used as a high-temperature gas cooled reactor nuclear control rod.

The invention relates to the field of reactor data processing, and particularly discloses a numerical calculation method for solving a neutron transport equation. The method comprises the following steps: 1, describing the geometry of a reactor core of a nuclear reactor by utilizing a construction entity geometric modeling method, and subdividing a curved surface grid along a curved surface structure; 2, directly expanding a neutron flux distribution function in each curved surface grid by using a multivariate polynomial; 3, deriving a neutron transport equation and an integral residual equation of a boundary condition under each curved surface grid; 4, solving geometric integral parameters of each curved surface grid; step 5, establishing an adaptive linear system; and step 6, iteratively calculating and solving a maximum characteristic value and an unknown expansion coefficient.

A molten salt nuclear reactor includes a vessel and a control blade assembly disposed within the vessel. The control blade assembly includes a drive rod and a control blade. The drive rod further includes an oblique surface wherein vertical motion of the drive rod and its oblique surface causes horizontal motion of the control blade into and out of a reactor section of the molten salt reactor.

An analysis method and apparatus for measurement uncertainty of power distribution of a reactor core, a device, a storage medium, and a product. The analysis method for measurement uncertainty of power distribution of a reactor core comprises: constructing a simulated operation state on the basis of an actual operation state of a reactor core, determining state parameters of the reactor core in the simulated operation state, and acquiring a group constant of the reactor core (S202); performing calculation on the basis of a material calculation subprogram and the state parameters to obtain material components of each fuel rod in the reactor core (S204); performing calculation on the basis of a power calculation subprogram and the material components of each fuel rod to obtain real power distribution of the reactor core, and acquiring detector reaction intensity of the reactor core (S206); on the basis of the state parameters and the group constant, determining power parameters of the reactor core in the simulated operation state, and performing flux graph simulation on the basis of the power parameters and the detector reaction intensity to obtain simulated power distribution of the reactor core (S208); and determining the measurement uncertainty of power distribution of the reactor core on the basis of a difference between the real power distribution and the simulated power distribution (S210).

One variation of a system includes: a nuclear reactor; a shield arranged about the nuclear reactor; a metallic coolant; and a set of melt seals. The nuclear reactor includes a pressure vessel and nuclear fuel arranged within a lower region of the pressure vessel. The metallic coolant includes a mixture of metals and is configured to: occupy a liquid state within an operating temperature range; occupy an interstitial volume between the nuclear reactor and the shield; and occupy the lower region of the pressure vessel encompassing the nuclear fuel. The set of melt seals are arranged on the pressure vessel and configured to open to enable transfer of a volume of the metallic coolant from the interstitial volume into the lower region of the pressure vessel in response to temperatures within the pressure vessel exceeding the operating temperature range.

The invention relates to the technical field of lead-based stacks, in particular to a nuclear storage integrated lead-based stack with a load autonomous following function. The nuclear storage integrated lead-based reactor with the automatic load following function comprises a reactor core, a phase change energy storage device and a heat energy utilization device, the reactor core is used for heating a coolant; the heat energy utilization device is used for absorbing heat in the coolant; the phase change energy storage device is arranged on the inlet side of the reactor core, the phase change energy storage device is used for carrying out heat interaction with a coolant, and the phase change temperature of the phase change energy storage device is consistent with the preset inlet temperature of the reactor core. According to the nuclear storage integrated lead-based reactor with the automatic load following function, natural circulation flow is not prone to oscillation and diffusion, the risk of overheating melting of the fuel assembly is not prone to occurring, thermal fatigue of the structural assembly is not prone to occurring, and high safety performance is achieved.

A mixed-spectrum molten-salt reactor has a primary containment vessel. A reactor core is housed in the primary containment vessel, the reactor core has an atypical geometry that holds a fuel salt. A blanket region is formed within the primary containment vessel and surrounding the reactor core, the blanket region comprises a blanket salt.

A method for controlling a pressurized water reactor, computer program product and control system, the pressurized water reactor includes a reactor core and a primary cooling circuit. The primary cooling circuit includes a primary cooling medium, which includes: acquiring a plurality of measurable reactor process variables and obtaining a plurality of non-measurable reactor process variables. The method further includes calculating future axial offsets at the end of a predetermined prediction time interval for a plurality of different possible boration / dilution actions based on the plurality of measurable reactor process variables and the plurality of non-measurable reactor process variables, the axial offset being a normalized difference between power of an upper half of the reactor core and a lower half of the reactor core. The calculation of the future axial offset for each of the plurality of different possible boration / dilution actions is performed in parallel.

The present invention relates to a nuclear fuel assembly suitable for a small modular reactor (SMR), the nuclear fuel assembly comprising: a spacer grid (110) having 17x17 grid cells; a plurality of fuel rods (120) arranged in the grid cells of the spacer grid (110); guide tubes (130) arranged and fixed in the grid cells of the spacer grid (110); and an upper end fitting (140) and a lower end fitting (150) fixed to the upper end and the lower end of the guide tubes (130), respectively, wherein the guide tubes (130) are characterized in that eight guide tubes are arranged in the main axis directions (Ixx)(Iyy) with respect to the central cell of the spacer grid (110), four guide tubes are arranged in the diagonal direction (Ixy), and sixteen guide tubes are arranged in the non-diagonal directions (Ixxy)(Ixyy).

A pool type reactor safety management system comprises a reactor shut-down device, a passive residual heat removal device, a main circulation device, an auxiliary circulation device and a passive circulation device, and the reactor shut-down device is used for realizing reactor shut-down under accident conditions; the passive residual heat removal device forms natural circulation of the reactor water pool and the external hot trap; the main circulating device provides a circulating pipeline and drives cooling water to circulate under a normal working condition; the auxiliary circulating device comprises an auxiliary circulating pump connected to the main circulating pump in parallel; and the passive circulating device comprises a natural circulating valve arranged in the lower cavity of the reactor. The system can effectively avoid overheating of the cooling water under the accident working condition, so that the flowing direction of the cooling water is quickly switched from forced convection under the normal working condition to natural convection under the accident working condition, the flowing stagnation is prevented, and the safety and the reliability of the pool type reactor are effectively improved. The invention also provides a pool type reactor safety management method.

The invention belongs to the technical field of processing of nuclear reactor control rod neutron absorber materials, and particularly relates to a preparation method of a nuclear-grade hafnium cast ingot. The method comprises the following steps: preparing sponge hafnium into an electrode block; performing vacuum electron beam melting on the electrode block for at least two times to obtain a finished cast ingot; and processing the finished cast ingot to prepare the nuclear-grade hafnium cast ingot which can be used for processing the control rod profile for the nuclear reactor assembly. According to the preparation method of the nuclear-grade hafnium cast ingot, the tensile strength, the welding performance and the creep performance of a nuclear-grade hafnium profile can be effectively improved; meanwhile, the technological process is simplified, and the technological economy and the production efficiency are remarkably improved; in addition, sponge hafnium is directly adopted for vacuum electron beam melting, so that the cost of raw materials is greatly reduced. According to the method, the application performance of the nuclear-grade hafnium profile is improved, meanwhile, the process is simplified, the production efficiency is improved, the production cost is reduced, and the economical efficiency of nuclear-grade hafnium profile manufacturing is greatly improved.

Provided are a hollow radioactive containment fuel containing a burnable poison, a preparation method therefor and the use thereof. The hollow radioactive containment fuel containing a burnable poison comprises a hollow fuel area and a fuel-free protective layer covering the periphery of the fuel area; the hollow fuel area comprises a matrix, and a fuel element and a burnable poison which are distributed in the matrix, wherein the fuel element is an isotropic coated fuel particle, and the burnable poison is an isotropic coated burnable poison particle. The present invention can realize highly compact and safe radioactive containment of a reactor fuel, ensure that radioactive products, especially fission gas, are not released into the environment accidentally, stably provide a fission site for the core of a miniature nuclear power pile for a long period of time, effectively confine the radioactive fission products inside the fuel, and guarantee the safe and stable operation of a reactor.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for relieving remote expansion of a cushion block locking cap of a heat insulation sleeve. Comprising a locking cap expansion device and a long rod, the long rod is used for remote operation above a pool, and the locking cap expansion device comprises a positioning pin, a positioning cylinder, a clamping gripper, an aligning and screwing operation head, an expansion head and a pressing expansion operation head. The device has the beneficial effects that a long-distance long rod is adopted to operate screwing positioning alignment and pressing expansion, and remote operation of locking of the cushion block locking cap is achieved. A special device integrating positioning of an expansion tool, screwing positioning of an expansion head, clamping and fixing and pressing expansion is adopted, underwater remote expansion integrated operation of the relieving cushion block locking cap is achieved, and the process steps are simplified.

The embodiment of the invention provides a reactor control method and device, electronic equipment and a storage medium, and belongs to the technical field of nuclear power. The method comprises the steps that the type of a steam turbine and a control strategy of a nuclear power unit are obtained, and the nuclear power unit comprises a reactor; according to the type of the steam turbine and the control strategy, an energy balance signal and a heat signal are calculated, the energy balance signal is used for indicating the minimum energy needed by operation of the steam turbine, and the heat signal is used for indicating the energy generated by the reactor; obtaining a feedforward signal according to the energy balance signal and the heat signal; and controlling the rod position action of the control rod according to the feedforward signal. According to the embodiment of the invention, the response speed and the adjustment capability of the nuclear power unit in the case of rapid load change can be improved.

To securely and easily disassemble and assemble a rack gear and a pinion gear. A drive mechanism 5 includes a rack gear 5A, a pinion gear 5B engaging with the rack gear 5A, a positioning member 5D fixed to a portion in the rack gear 5A where rack teeth 5Ab are not included, and a plurality of abutting members 5E disposed along a rotation direction of the pinion gear 5B. In a case in which at least one of the rack gear 5A and the pinion gear 5B is moved in a relative movement direction at the time when the rack gear 5A engages with the pinion gear 5B, abutment between the positioning member 5D and the abutting member 5E defines a rotational position of pinion teeth 5Bb of the pinion gear 5B to be an engaging position with the rack teeth 5Ab of the rack gear 5A.

A method includes: during a first time period, pumping propellant: to a set of moderator coolant channels to cool the moderator; through a set of fuel elements to heat the propellant to a first outlet temperature; out of a thrust nozzle to produce a first thrust; and locating the reflector at a first closed position to reflect neutrons toward the set of fuel elements to maintain a fission rate of the set of fuel elements above a threshold fission rate. Additionally, during a second time period: locating the reflector at a second open position to reduce the fission rate of the set of fuel elements to below the threshold fission rate; and pumping the propellant through the fuel elements to heat the propellant to a second outlet temperature less than the first outlet temperature and out of the thrust nozzle to produce a second thrust less than the first thrust.

The embodiment of the invention provides a self-adjusting shielding heat preservation device for a micro reactor and an operation method of the self-adjusting shielding heat preservation device, and relates to the field of nuclear reactors. The reactor comprises an outer shell, an inner shell, a control rod and a spiral flow channel, the inner shell is sleeved with the outer shell, an outer air inlet hole and an outer air outlet hole are formed in the outer shell, an inner air inlet hole and an inner air outlet hole are formed in the inner shell, a reaction cavity is formed in the inner side of the inner shell, and the spiral flow channels are arranged on the outer side of the inner shell and the inner side of the outer shell. The control rod is rotatably arranged on the side wall of the inner shell in a penetrating mode and located on the inner side of the outer shell, the control rod is provided with a control area and a non-control area, and the spiral runner is arranged between the inner shell and the outer shell; the outer gas inlet hole is used for inputting cooling gas, so that the cooling gas sequentially passes through the inner gas inlet hole, the reaction chamber, the inner gas outlet hole and the spiral flow channel and then is discharged from the outer gas outlet hole, and the cooling gas is used for flowing along the spiral flow channel and enabling the control rod to rotate so as to drive the control region to rotate into or out of the reaction chamber.

The invention discloses a research reactor fault detection method, device and system and a storage medium, relates to the technical field of nuclear power, and aims to at least solve the problems that a fault detection result is fuzzy and inaccurate, the troubleshooting workload is large and the fault processing efficiency is low. The method is applied to a research heap system and comprises event nodes, state nodes and a preset change trend, caused by fault events corresponding to the event nodes, of the state nodes associated with the event nodes. Comprising the steps of determining a first event node associated with a target state node under the condition that the target state node is detected to be in a fault state; determining a first state variable quantity of a state variable of a first state node associated with the first event node; and selecting a second event node of which the change trend of the first state change quantity associated with the event is the same as the target preset change trend corresponding to the first state node from the first event nodes, so as to determine the fault event of the second event node as a target fault event.

A fast neutron nuclear reactor contains a nuclear reactor core having an array of device locations. Some device locations in the nuclear reactor core contain fissile and fertile nuclear fuel assembly devices. One or more other device locations in the nuclear reactor core contain Doppler reactivity augmentation devices that amplify the negativity of the Doppler reactivity coefficient within the nuclear reactor core. In some implementations, a Doppler reactivity augmentation device can also reduce the coolant temperature coefficient within the nuclear reactor core. Accordingly, a Doppler reactivity augmentation device contributes to a more stable nuclear reactor core.

The present invention relates to a method for producing a nuclear design core model accounting for load following, the method comprising the steps of: receiving load-following requirements for a corresponding cycle, wherein the load-following requirements include the amount of power reduction, the number of load-following operations, and at least one of the time or duration of operation at reduced power; determining the amount of insertion of a control rod on the basis of the load-following requirements; determining the duration of insertion of the control rod on the basis of the amount of insertion of the control rod and the maximum insertion limit thereof; determining the time of insertion of the control rod by simulating core combustion by reflecting the duration of insertion of the control rod in a nuclear design code at each of a plurality of time points of the corresponding cycle; and producing a nuclear design core model accounting for load following on the basis of the determined amount of insertion of the control rod, the determined duration of insertion of the control rod, and the determined time of insertion of the control rod.

A nuclear reactor shutdown system includes a shielded path that passes through a reactor core fuel housed in a nuclear reactor vessel in a hermetically sealed manner, one end of the shielded path having an opening and another end of the shielded path being closed; a neutron absorber that is allowed to enter from the opening of the shielded path; an elastic member that is configured to urge the neutron absorber in a direction entering inside from the opening of the shielded path by being released from a compressed state; and a braking part that is disposed so as to maintain the compressed state of the elastic member and is configured to release the elastic member from the compressed state when the braking part reaches or exceeds a threshold temperature.

The invention discloses a control rod driving mechanism of a small-sized modularized pressurized water reactor, which belongs to the technical field of nuclear, and comprises a protective shell with the radial size of not more than 70mm; a control rod driving main body is mounted in the protective shell, the control rod driving main body comprises a flange which is located at the left end of the protective shell and used for being connected with a reactor cover plate, a precise driving mechanism is arranged on the right side of the flange, and a gear box is mounted on the right side of the precise driving mechanism; a position sensor unit is installed on the right side of the gearbox, and a sealed electrical input end is arranged on the right side of the position sensor unit. The control rod driving mechanism of the small modular pressurized water reactor is compact in size, the stroke exceeds 1.5 m, the movement increment of each time is smaller than 5 mm, it is ensured that operation of a reactor core of the reactor is accurately controlled, and meanwhile the control rod driving mechanism can stably operate under the conditions of high temperature, high pressure, ionizing radiation and corrosion.

The application relates to a reactor core power distribution measurement uncertainty analysis method, device, equipment, storage medium and product. The method comprises the following steps: constructing a simulation operation state based on an actual operation state of a reactor core, determining state parameters of the reactor core in the simulation operation state, and obtaining group constants of the reactor core; performing calculation based on a material calculation subprogram and the state parameters to obtain material compositions of each fuel rod in the reactor core; performing calculation based on a power calculation subprogram and the material compositions of each fuel rod to obtain a real power distribution of the reactor core and obtain detector reaction intensities of the reactor core; determining power parameters of the reactor core in the simulation operation state based on the state parameters and the group constants, and performing flux map simulation based on the power parameters and the detector reaction intensities to obtain a simulation power distribution of the reactor core; and determining a measurement uncertainty of the reactor core power distribution based on a difference between the real power distribution and the simulation power distribution. The method can improve the accuracy and acquisition speed of the reactor core power distribution measurement uncertainty.

The invention provides a multipurpose small reactor control method, device, equipment and medium, and the method comprises the following steps: in an automatic temperature control mode, if the deviation between the actually measured temperature of a reactor outlet and the set temperature meets a preset automatic temperature control condition, determining a first control instruction of each adjusting rod of an adjusting rod group; a first control instruction is sent to an adjusting rod set executing mechanism so as to control an adjusting rod to move till the deviation between the actually measured temperature and the set temperature is within the temperature adjusting dead zone range; in the power automatic control mode, if the deviation between the set power and the actually measured power of the reactor meets a preset power automatic control condition, determining a second control instruction of the circulating pump of the intermediate loop; and a second control instruction is sent to the circulating pump executing mechanism so as to control the rotating speed of the circulating pump until the deviation between the set power and the actually measured power is within the power adjusting dead zone range. According to the invention, the temperature and the power are respectively controlled, and the control performance is optimized.

A control rod drive mechanism having a torque tube, a control rod assembly including a connecting rod and a spider, a lock cam assembly rotatably secured to a bottom end of the connecting rod and including a locking cam, and a locking collar disposed non-rotatably within the spider, the locking collar including a locking recess with an entry slot, wherein in a first axial position the lock cam assembly is rotatable with respect to the torque tube, and a second axial position the lock cam assembly is non-rotatable with respect to the torque tube.

A reactor control system (100) for a nuclear fission reactor system comprising a nuclear reactor unit (300) having a reactor core (302) with a central axis (320) extending along the length of the reactor core (302). The nuclear reactor unit (300) comprises a first region (310) provided in the reactor core (302) and a second region (312) provided outside of the reactor core (302). The control system (100) comprises a control duct (200) having a first portion (210) configured to extend through the first region (310) and a second portion (220) configured to extend through the second region (312). The duct (200) is filled with a series of control units (400), each of the control units (400) configured to travel along the control duct (200).

The present invention relates to a reactor construction comprising an upper compartment above a reactor compartment, said reactor compartment being separated from the upper compartment by a radiation shield, wherein a molten salt reactor (MSR) is situated in the reactor compartment, the MSR comprising a reactor vessel with a reactor vessel lid assembly comprising a reactor vessel lid shield and a reactor vessel lid wherein a lifting element is situated in the reactor construction, the lifting element comprising: at least one bar being substantially vertical and axially adjustable along its longitudinal axis by a drive mechanism situated in the upper compartment, and wherein at least one through-hole for the at least one bar is provided in the radiation shield and wherein the reactor vessel lid shield has at least one receiving structure configured to engage with and lock to the at least one bar. The invention also relates to a method of performing a maintenance operation in the reactor compartment and a power barge comprising at least one reactor construction.

The application relates to a neutron absorption material and application thereof to a control rod for a nuclear reactor core. The neutron absorption material provided by the application is prepared by mixing ytterbium elements and holmium elements in a certain proportion, can effectively play the high neutron capture cross-section characteristics of the ytterbium elements and the holmium elements, effectively absorb thermal neutrons, and can ensure good chemical and thermal stability, so that the neutron absorption material can maintain its structure and performance stability for a long time in a high irradiation environment, effectively avoids the conversion of the neutron absorption material into a weak neutron absorption material after absorbing neutrons, and further controls the neutron absorption characteristics of the neutron absorption material. Further, the neutron absorption material provided by the application can still ensure the neutron absorption characteristics thereof and maintain the reactivity value under high burnup depth, so as to prolong the service life of the control rod for the nuclear reactor core.

The invention discloses a nuclear reactor protection system with built-in diversity. A system architecture with three sequences, local coincidence logic, two-stage two-out-of-three voting and parallel processing of two technical routes is adopted. Wherein the three sequences are three multiple sequences, each sequence is composed of a plurality of subsequences adopting different technical route devices, and three subsequences adopting the same technology form a subgroup; each subgroup has the capability of independently executing a security function; system architecture configuration with built-in diversity is adopted, additional arrangement of a DAS system is not needed, better compactness is achieved, and the requirement of a small reactor for space is better met; a plurality of subgroups use chips based on different technologies, adopt different architecture schemes, different development tools and logic design methods, have independent safety function driving paths, and adopt diversified means for verification and confirmation, so that common cause faults can be effectively handled.