299results about "Plant parameters regulation" patented technology

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

The invention provides a start-stop auxiliary device used in an integral reactor and a cold starting method of the integral reactor. An inlet of a starting separator is connected with an outlet conduit of a once-through steam generator, an outlet of the starting separator is connected with a main steam pipe through a steam valve, and a drain outlet is connected with a water storage tank. An outlet at the bottom of the water storage tank is connected with a circulating water pump, and one conduit at the outlet of the water storage tank is connected with a condenser through an outlet stop valve of the water storage tank, a water storage tank water level control valve and a thermoreduction and dropping device. The circulating water pump is connected with a water supply conduit at the inlet of the once-through steam generator through a circulating water-carrying control valve. A water feed pump is connected with the condenser, the water feed pump is further connected with the inlet conduit of the once-through steam generator through a main water feed valve and with the water storage tank through a cold water valve. The start-stop auxiliary device is capable of increasing the temperature and pressure of refrigerant in a first circuit when the reactor starts, starting pressure and starting flow of a second circuit are established effectively, phenomenon of unstable flow is avoided, working media and heat energy are recycled effectively, and starting time is shortened.

The object of this invention is to provide a method for mitigating a stress corrosion cracking of reactor structural material which makes it possible to suppress the rise in the main steam line dose rate without secondary effects such as a rise in the concentration of radioactive cobalt-60, etc. in the reactor water. Hydrogen and a reductive nitrogen compound containing nitrogen having a negative oxidation number (for example, hydrazine) are injected into the core water of boiling water nuclear power plant. By injecting the reductive nitrogen compound containing nitrogen having a negative oxidation number into the core water, the stress corrosion cracking of structural material of reactor can be mitigated without side reactions such as a rise in the concentration of cobalt-60, etc.

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

The invention discloses a nuclear power plant steam turbine valve fault diagnosis method, an auxiliary diagnosis method thereof, and a test device. Through a user instruction corresponding to preset parameter information, the control of state switching action of a valve is carried out, and when the valve acts, the parameter information of switching action of the valve is recorded. Then whether the recorded parameter information and the preset parameter information are matched or not is judged, fault information is generated if not, thus the online fault diagnosis of the steam turbine valve of a nuclear power plant is realized, through comparing the various parameter information generated in the action of the valve and the preset parameter information, the accuracy of judging a problem in a diagnose or test is improved, the applicability is strong, the fault information of most valves in action can be diagnosed, and the efficiency of diagnosis is improved.

A recombination device (1, 1′) for catalytically recombining hydrogen and/or carbon monoxide with oxygen in a gaseous mixture comprises at least one catalyst system (2) in which a housing (4) is mounted through which the gaseous mixture can flow in free convection in the operational phase. According to the invention, said catalyst system (2) is provided with a plurality of sub-areas (T1, T2) in the direction of flow. A first sub-area (T1) comprises in the incoming direction a catalyst body (6) with a surrounding throttle layer (8) for inhibiting the diffusion of the incoming and/or discharged reaction gases. A second sub-area (T2) that adjoins the first sub-area (T1) comprises at least one catalyst body (6) that is directly accessible by the reaction gases.

The invention discloses a debugging method of a nuclear power station steam turbine control system. The method comprises the steps that a nuclear power station steam turbine control system simulation model is established, the simulation model at least comprises a simulation debugging device; the nuclear power station steam turbine control system is connected with the simulation model, the nuclear power station steam turbine control system at least comprises one of a steam turbine adjusting subsystem, a steam turbine protecting subsystem, a steam turbine monitoring subsystem (GME) and a steam turbine shaft seal subsystem, each subsystem is subjected to simulation debugging through the simulation debugging device, and according to the simulation debugging results, the functions of the subsystems are verified. According to the debugging method of the nuclear power station steam turbine control system, the simulation model is used for being connected with the steam turbine control system, system function verifying is carried out in advance, nuclear power plant debugging scheme optimizing is achieved, the debugging period is shortened, and the risk of steam turbine unexpected halt and the risk of reactor unexpected tripping after a nuclear power unit is started are lowered. In addition, the invention further discloses a debugging device and system of the nuclear power station steam turbine control system.

The invention discloses a diversified driving method for a nuclear power station. The method comprises the following steps: acquiring parameter information of each process in operation of the nuclear power station; comparing the parameter information of each process with a preset corresponding set value; outputting a signal for driving the execution of automatic functions if more than one but not all of the process parameter information exceeds the corresponding set value, wherein the automatic functions include an automatic tripping function of a steam turbine, an automatic start function of an auxiliary water supply electric pump, an automatic start function of a steam driven pump, a main steam automatic isolation function, a main water supply automatic isolation function and an automatic pump shutdown function of a reactor coolant pump. Besides, the invention also discloses a diversified driving device and a diversified driving system for the nuclear power station.

A system and method is provided for a continual updating of the optimization of multiple operational control-variables during the operation of a nuclear reactor over a plurality of fuel cycles. A networked computer system includes one or more hosts programmed to execute an optimization process to identify and make changes in quantitative values of operational control-variables that result in improved efficiency and operational flexibility. Optimization and updating of operational control-variables may proceed selectively under manual control for inputting specific optimization constraints and reactor state-point information or may proceed autonomously through a repetitive performing of the optimization process based upon a predetermined user-defined strategy stored on the network. Communications between users and networked processors is facilitated by use of a TCP/IP server connected to the Internet so that portions of the optimization process may be conducted contemporaneously at remote locations and/or the results made accessible to users via conventional browser enabled computers.

The invention is applicable to the professional technical field of mega-kilowatt-class pressurized water reactor nuclear power stations and provides a pressurized water reactor nuclear power station steam generator water level control method and system. The method includes a steam engine bypass drain valve openness signal, a water feeding deaerator steam inlet valve openness signal and a steam engine steam inlet narrow range signal; adopting a preset secondary loop image load algorithm to acquire secondary loop image load according to the steam engine bypass drain valve openness signal, the water feeding deaerator steam inlet valve openness signal and the steam engine steam inlet narrow range signal; acquiring a water level control signal; adopting the water level control signal to modify the acquired secondary loop image load and adopting the modified secondary loop image load to control a bypass water feeding adjusting valve of the steam generator. By means of the method and system, the steam generator water level out-of-control problem is solved, and heap jumping caused by the fact that the steam generator water level is out of control is avoided.

The invention relates to a method for regulation of operational parameters of the core of a pressurised water nuclear reactor comprising: a step of acquisition of values (FH, FB, TBC, TBF, Q) which are representative of the conditions of operating of the core of the reactor; a step of evaluation of the actual values (Tmoye, AOe, {circumflex over (P)} maxe) of the operational parameters at least according to the values acquired (FH, FB, TBC, TBF, Q);

• • a step of selection of a control law for the concentration of a neutron-absorbent component ([B]) and for the positions of insertion (Z1 to Z5) of the groups (P1 to P5) of rods selected from at least the first and second control laws which are different from one another; and a step of regulation of the operational parameters by means of the control law selected, according to set points (Tmoyc, AOc, {circumflex over (P)} maxc) relative to the said parameters and to the actual values (Tmoye, AOe, {circumflex over (P)} maxe) evaluated.

Methods for evaluating robustness of solutions to constraint problems include: (a) determining one or more inputs for the constraint problem; (b) determining bias, uncertainty, or bias and uncertainty of the one or more inputs; (c) randomly perturbating at least one of the one or more inputs based on the bias, uncertainty, or bias and uncertainty of the one or more inputs to determine one or more modified inputs; (d) generating a solution to the constraint problem based on the one or more modified inputs; (e) storing the solution; (f) repeating (c) through (e) until multiple solutions are stored; and/or (g) determining the robustness of the stored multiple solutions by comparison of the stored multiple solutions to each other. The one or more modified inputs include the randomly perturbated input or inputs and/or the non-perturbated input or inputs.

The invention discloses a debugging method and system for a closed-loop control system of a nuclear power plant. The debugging method comprises the following steps: A, a parameter design value input in an off-line checking table by a user is received to generate a checking message; B, a control parameter of a PID (Proportion Integration Differentiation) controller is verified and adjusted according to the checking message; C, the verified and adjusted control parameter is optimized under a disturbance operating condition; D, the control parameter of the PID controller and/or the logic and related constant value of the PID controller are/is modified under an instantaneous condition so as to ensure the normal response of the closed-loop control system under the instantaneous condition. According to the technical scheme disclosed by the invention, the debugging risk of the closed-loop control system is reduced, the debugging cycle of the closed-loop control system is shortened, the operation quality of a unit is improved, and good economic benefits are obtained.

The invention discloses control modes for a PWR core of a nuclear power plant. Two control rod groups, i.e., a T rod group and a D rod group, which are controlled independently are arranged. In the load tracking mode, a reactor is controlled in the manner that the average temperature of a coolant is controlled by the independent T rod group, reactive change of the core is controlled, the axial power distribution shape is controlled by the independent D rod group, soluble boron is adjusted periodically, and the control rod groups are recovered to an operation range. In other modes except the load tracking mode, the reactor is controlled in the manner that the average temperature of a coolant is controlled by the independent D rod group, the D rod group is also used to adjust the axial power shape and slight reactive change of the core. The control modes of the invention have the advantages that the control rod is used to reactive control of the core in the load tracking mode, so that burden of operators is greatly reduced; and generated wasted boron solution is reduced, the operation cost as well as the wasted solution processing cost is reduced, and the environment is protected.

A nuclear power plant is provided including a BWR, a reactor cooling system cooling the BWR, an HWC hydrogen injection system connected to the reactor cooling system and an alcohol injection system connected to the reactor cooling system. Methods for providing methanol and hydrogen are also provided.

In a method of determining an operating margin to a given operating limit in a nuclear reactor, operational plant data from an on-line nuclear reactor plant is accessed, and reactor operation is simulated off-line using the operational plant data to generate predicted dependent variable data representative of the given operating limit. The predicted dependent variable data is normalized for evaluation with normalized historical dependent variable data from stored operating cycles of plants having a similar plant configuration to the on-line plant. A time-dependent average bias and a time-dependent uncertainty value for the predicted dependent variable data are determined using the normalized historical dependent variable data, and a risk-tolerance level for the on-line plant is obtained. An operating margin to the given operating limit is determined based on the determined time time-dependent average bias value and time-dependent uncertainty value so as to satisfy the risk-tolerance level of the evaluated plant.

The invention discloses a system and a method for verifying functions of a starting and stopping reactor of a high-temperature gas cooled reactor. A secondary side inlet of a steam generator is communicated with an outlet of a secondary circuit main water supply system; an outlet of an auxiliary boiler is divided into two paths, wherein one path is communicated with an inlet of a steam superheater; the other path, an outlet of the steam superheater and an outlet of the steam generator are subjected to pipe combining through a pipeline and then divided into two paths, wherein one path is communicated with an inlet of a steam separator; the other path is communicated with an inlet of a main steam by-pass; a steam side outlet of the steam separator is communicated with an inlet of the main steam by-pass; a water draining side outlet of the steam separator is communicated with an inlet of a condenser; and an outlet of the main steam by-pass is communicated with the condenser. The system and the method can be used for verifying the functions of the starting and stopping reactor before a unit is started and operated, thereby guaranteeing the reliability and the safety in the process of starting and stopping a high-temperature gas cooled reactor unit.