534 results about "Pressurized water reactor" patented technology

The invention provides a pressurized water reactor scale-caused axial power offset risk coupling calculation method and device. The method comprises the following steps: performing corrosion scale thickness deposition simulation by using a first calculation module, and calculating to obtain cooling liquid temperature distribution in the axial direction of a fuel rod and corrosion scale deposition thickness distribution in the axial direction of the fuel rod; a second calculation module is used for carrying out enrichment simulation on boron elements in the corrosion scale deposits and calculating the boron concentration distribution condition in the corrosion scale deposits of each layer in the axial direction; and performing fuel rod axial power distribution simulation by using a third calculation module, and calculating to obtain a scale-caused axial power offset value according to the corrosion scale thickness distribution and the boron concentration distribution condition. The scale-caused axial power offset value is predicted by adopting a multi-module, multi-physical-field and multi-scale coupled calculation method, so that the technical problem of how to comprehensively consider the effects of multiple scales and multiple physical fields to predict and analyze corrosion scale deposition is solved.

The invention discloses a nuclear reactor core channel analysis method based on two fluid models, which comprises the following steps of: 1, establishing a fluid domain computational grid according to the structure of a nuclear reactor core or an assembly; 2, respectively establishing a control equation set and a heat conduction model of the fuel rod for the continuous gas-liquid phase and the discrete liquid drop phase; 3, considering physical phenomena such as flow heat exchange, resistance pressure drop and turbulence mixing in a reactor core, and implanting a corresponding constitutive model; 4, discretizing the control equation based on a second-order discrete format to obtain a discrete equation set; 5, dividing parallel tasks and parallel regions through an MPI parallel strategy to form a large coefficient matrix of a pressure equation; and 6, solving a discrete equation set and a large coefficient matrix through a pressure coupling semi-implicit algorithm and a stable biconjugate gradient method. According to the method, the key physical parameter distribution of the pressurized water reactor can be more accurately and reasonably calculated, so that technical guidance is provided for safety analysis and engineering design of the reactor.

The invention relates to the field of nuclear power unit accident analysis, in particular to a coupling method of a VVER unit system program and a neutron dynamics program. The method comprises the following steps: adding an interface function in a reactor core three-dimensional neutron dynamics program, compiling the reactor core three-dimensional neutron dynamics program into a dynamic link library, and accessing and calling the dynamic link library by a reactor system program in a plug-in manner; establishing a neutronics physical model of a specific loading scheme of the VVER type reactor; establishing an accident analysis system simulation model of the VVER type pressurized water reactor; establishing control logic of main systems and equipment of a primary loop and a secondary loop in the VVER type reactor; establishing control logic and conservative condition settings for specific accidents; executing coupling calculation of a reactor system program and reactor core three-dimensional neutron dynamics; and obtaining key safety parameters of the VVER type reactor in a specific accident based on a coupling calculation result. According to the invention, simulation of changes of physical and thermal parameters in a primary loop system and a secondary loop system and action response of a main system after a reactor has an accident can be realized, and efficient and reliable accident analysis can be conveniently carried out on a VVER type nuclear power unit.

The invention discloses a pressurized water reactor rod changing method and system, electronic equipment and a computer readable storage medium, and the method comprises the steps: obtaining a plurality of rod pre-changing strategies corresponding to a first burnup point according to a plurality of rod changing interpolation relationships of a plurality of control rods when it is monitored that the current burnup point is the first burnup point, each control rod corresponds to a rod changing interpolation relation, the rod changing interpolation relation comprises a plurality of burnup points and a rod pre-changing strategy corresponding to each burnup point, and the rod pre-changing strategy is used for indicating the pre-insertion step number of the corresponding control rod; adjusting the pre-insertion step number of each control rod according to a rod pre-changing strategy corresponding to each control rod at the first burn-up point; and rod replacement processing is carried out based on the multiple adjusted control rods. According to the invention, disturbance of the pressurized water reactor can be reduced when rod sequence exchange is carried out on the control rod.

The invention discloses an in-reactor fuel irradiation test loop based on the working condition of a pressurized water reactor. The in-reactor fuel irradiation test loop comprises a main circulation system, a purification system, a fission substance release monitoring system, a water supplementing system, an emergency cooling system, a cooling water system, an in-reactor device pressure failure online monitoring system, a reactor shutdown protection system, a thermal parameter measurement system, a power supply system and an air ventilation and exhaust system. According to the loop, a pressurized water reactor working condition environment is realized, and various technical indexes such as temperature, pressure, flow velocity and water quality required by fuel test are ensured to be reached; the safety level of each system is reasonably configured, so that the requirements of laws and regulations are met, and the safety performance of the loop is improved; by reasonably arranging a multi-stage cooling water system, an air exhaust system and the like, the radioactive activity of an exhaust flow is reduced, the environmental safety is ensured, the release of radioactive aerosol in a process is eliminated, the exposure dose of workers is reduced, the safety of the workers is ensured, the safety risk of devices in the reactor to the reactor under the working condition of a fracture accident is eliminated, and the reliability of the system is improved.

The invention discloses a neural network-based axial power offset prediction method for a peak regulation process of a pressurized water reactor, and the method comprises the steps: 1, arranging power, fuel enrichment degree, the number of burnable poison rods in a fuel assembly, control rod positions, burnup level, axial xenon poison distribution, axial samarium poison distribution, radial xenon poison distribution, radial samarium poison distribution and the like according to a specific format; and making a standard input file. Secondly, arranging the AO (axial power offset) value of the reactor core under each target point in a specific format, and making a standard output file; thirdly, building a convolutional neural network model; and 4, putting the standard input file and the standard output file into the model for training until model parameters meet requirements, and exporting the trained model parameters. And 5, making a standard input file of a target point needing to be predicted according to the method in the first step. Sixthly, the standard input file of the target points needing to be predicted is calculated by loading model parameters, and the AO predicted value of each target point is obtained.

A pressurized water reactor oxidation corrosion product cross-scale nuclear-thermal-material coupling simulation method comprises the following steps: in the first stage, obtaining a supercooled boiling position through mutual loop iteration simulation of a neutronics model and a thermal hydraulic model; and in the second stage, multi-physics field coupling simulation is carried out through an oxidation corrosion product (CRUD) model, a neutronics model and a thermal hydraulic model, and high-resolution simulation under the pressurized water reactor nuclear-thermal-material coupling condition is obtained. According to the method, a scene in which the thickness range of the oxidation corrosion product is 20-200 microns is considered, multi-physical fields such as pressurized water reactor neutron physics, flow heat transfer and oxidation corrosion product deposition are calculated and analyzed, and the complex physical process can be captured more accurately.

The invention relates to a pressurized water reactor MOX fuel powder preparation method, which comprises: preparing a mixed actinide eutectic melt from a mixed actinide salt and a urea solid, and carrying out deamination, denitrification / decarbonization and reduction on the mixed actinide eutectic melt to directly obtain MOX fuel powder, the mixed actinide salt is hexavalent actinide nitrate and tetravalent actinide oxalate; when the mixed actinide salt is a uranyl nitrate hexahydrate solid and a thorium oxalate hexahydrate solid, the MOX fuel powder is (UxTh1-x) O2; when the mixed actinide salt is a uranyl nitrate hexahydrate solid and a plutonium oxalate hexahydrate solid, the MOX fuel powder is (UxPu1-x) O2. According to the method provided by the invention, the raw materials are simple, the preparation process flow of the MOX fuel powder composed of UO2 and PuO2 (or ThO2) can be greatly simplified, the use of process reagents is reduced, and no process waste liquid is generated, so that the MOX fuel powder preparation method which is simple, convenient, high in production efficiency, free of waste liquid and more economical and green is provided.

The invention discloses a system for LOCA accident simulation irradiation test and a design method, which are characterized in that a pressurized water reactor fuel element LOCA irradiation performance test is taken as a research object, a circulation loop and a test device are established, prototype reactor operation conditions and different accident process thermal hydraulic conditions are simulated, a fuel element LOCA irradiation simulation test is carried out, and a test result is obtained. The high-temperature and high-pressure water system is divided into three areas according to temperature and pressure, the three-stage cooling circulation system is combined with the detachable plate heat exchanger, and the number of the heat exchangers and the parameter coupling complexity among the three-stage system are reduced. The mode of combining circulation loop electric heating and in-pile device electric heating is adopted, the irradiation test thermal condition is controlled, the test power of the device is increased, the device adopts a long heat exchange flow channel, the natural circulation capacity is constructed, and the safety of the test device is improved; the invention provides a test method aiming at release working conditions of trace and massive gases and fission products, and a simulated irradiation test covering various accident working condition ranges is provided, so that the application range is wide.

The invention belongs to the technical field of nuclear power, and particularly relates to a multi-scale coupling determination method and device for the deposition process of dirt on the surface of a fuel rod in a pressurized water reactor. According to the method, a multi-scale coupling solution scheme is adopted, fluid turbulence energy and temperature distribution obtained through macro-scale solution are used for solving a sedimentary layer internal temperature and pressure control equation on the micro-mesoscale, and the deposition rate is obtained; the required material input parameters are less and can be adjusted in real time along with the operation state of the pressurized water reactor; besides, parameters such as surface temperature, pressure and the like of the sedimentary layer obtained by micro-mesoscale solution can be used as boundary conditions to solve a macroscopic flow equation, so that a bidirectional coupling effect is realized, the application range of the method is expanded, and the calculation accuracy is improved. The parameter setting of the method disclosed by the invention can adopt a physical property change function of the material along with the change of temperature, pressure and the like, so that the universality and practicability of a program are improved.

The invention discloses a pressurized water reactor primary loop boron concentration prediction and control method and system, and the method comprises the steps: obtaining an original helium flow sequence of a pressurized water reactor primary loop, and obtaining M boron concentration IMF components IMF1,..., IMFM; the M boron concentration IMF components IMF1,..., IMFM are put into a trained LSTM model for prediction, predicted values Y1, Y2,..., YM of the IMF components are obtained, and error coefficients delta Y1, delta Y2,..., delta YM corresponding to the boron concentration IMF components are calculated according to the predicted values Y1, Y2,..., YM of the IMF components; and inputting error coefficients delta Y1, delta Y2,..., delta YM corresponding to each boron concentration IMF component into the trained DenseNet prediction model to obtain an optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence, based on a DEM-DAC algorithm, controlling the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs1 # of the boron concentration signal sequence, and calculating the boron concentration of the pressurized water reactor primary loop according to the optimal IMF component predicted value # imgabs0 # of the boron concentration signal sequence. The method and the system can solve the technical problems that the boron concentration fluctuation of the primary loop is large and the unit load regulation performance is poor during the reactor variable load operation period.

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.

A pressurized water reactor core oxidation corrosion product deposition multi-physical coupling hardware-in-loop device comprises a corrosion product accelerated deposition system, an element online testing system and an axial power offset digital mirror image system. According to the method, the complex multi-physics field coupling process of hydrochemistry, material corrosion, flow heat transfer, neutron physics and the like involved in oxidation corrosion product deposition can be achieved, off-line data transmission between single physics fields is avoided, the growth rule of a deposition layer is directly obtained, the deposition mechanism of oxidation corrosion products is analyzed, and the deposition efficiency is improved. The defects that an existing device cannot perform entity verification and cannot obtain element concentration and distribution in a corrosion product deposition layer are overcome.

The invention provides a mass evaporation rate calculation method and device of a boiling environment and a readable medium. The mass evaporation rate calculation method comprises the following steps: carrying out node division on a pressurized water reactor coolant system; for each node, whether the dirt thickness of the current node is larger than a critical value or not is judged, if yes, the current node is the dirt surface, the mass evaporation rate of the node is calculated based on a porous sediment boiling module, and if not, the current node is the clean cladding surface, and the mass evaporation rate of the node is calculated based on the boiling heat flow and the mass evaporation module. According to the method, clean and dirty cladding surfaces are respectively considered, a group of mass evaporation rate calculation formulas are established, the mass evaporation rates of the reactor core and each node of the steam generator are calculated in detail according to information input by a user, and the calculated mass evaporation rates are more accurate and accord with actual scenes.

The invention discloses a method for improving the power load following flexibility of a pressurized water reactor, which comprises the following steps: arranging a bypass branch line on a steam bypass pipeline on the secondary side of a steam generator, and arranging a heat energy storage device on the bypass branch line; steam on the secondary side of the steam generator is shunted into the heat energy storage device in proportion through a bypass branch line; when the power load is low, secondary side steam of the steam generator is introduced into the heat energy storage device for heat storage; when the power load is high, the heat energy storage device releases stored heat and heats condensed water, and generated steam is output from the heat energy storage device to be used for power generation. The method does not reduce the operation power of the nuclear reactor, does not cause energy waste, can cope with the mismatching between the power supply and the power demand, maintains reliable energy distribution, and improves the power load following flexibility of the pressurized water reactor.

The invention discloses a reactor core for achieving 18-month circulation of a pressurized water reactor with the power exceeding 4000 MW and a loading method of the reactor core, the reactor core is composed of 177 groups of annular fuel assemblies, a 20-month refueling strategy is adopted in the first circulation, each assembly is divided into four areas to be loaded, the enrichment degrees of the assemblies in the four areas are 4.1%, 4.25%, 4.45% and 4.95% respectively, the number of the assemblies is 21, 64, 72 and 20 groups respectively, and the number of the assemblies in the four areas is 4.1%, 4.25%, 4.45% and 4.95% respectively. Placing 4.95% of the components in an outer area of the reactor core, and arranging 4.45% of the components, 4.25% of the components and 4.1% of the components in an inner area of the reactor core according to a checkerboard format; in the second cycle and the third cycle, a low-leakage refueling mode is adopted, and 2 / 3 of the reactor core, namely 108 groups of new assemblies, is loaded into the reactor core during refueling each time; and in the fourth cycle, balanced refueling is achieved, and the reactor core adopts a low-leakage refueling mode. According to the invention, after the annular fuel assembly is adopted by the pressurized water reactor consisting of 177 groups of fuel assemblies, the reactor core power is improved by 30% and reaches 4130MW compared with that when rod-shaped fuel is adopted, and the balance cycle life length still meets 18 months, so that the economical efficiency of the reactor core is obviously improved.

The invention provides a pressurized water reactor core axial power offset phenomenon prediction method and electronic equipment. The prediction method comprises the following steps: firstly, constructing a prediction model of the axial power offset phenomenon of the reactor core; wherein the prediction model is used for representing a model that the axial power offset phenomenon of the reactor core changes along with oxidation corrosion sediment CRUD and fuel burn-up on the surface of a cladding, and the prediction model comprises a neutron physical module, a thermal hydraulic module and a CRUD material module which have a coupling relationship; and then, carrying out coupling calculation on the prediction model within a preset duration, and generating prediction information of the axial power offset phenomenon of the reactor core. According to the prediction mode, the reactor core axial power offset phenomenon changing along with fuel burnup and CRUD deposition in the operation process of the pressurized water reactor is described, and therefore the prediction precision of the reactor core axial power offset phenomenon is improved.

The invention discloses a reactor core heat exchange test device, which comprises a test body, the test body comprises an internal component and an external component, the internal component comprises a flow channel and a reactor core rod bundle, the reactor core rod bundle is arranged in the flow channel and extends along the extension direction of the flow channel, the reactor core rod bundle is a heating rod, and the external component is arranged in the flow channel and extends along the extension direction of the flow channel. The external component comprises an upper shell, a reactor core shell and a lower shell which are connected in sequence, the upper shell is provided with an upper cavity, the lower shell is provided with a lower cavity, an inner cavity of the reactor core shell is communicated with the upper cavity and the lower cavity, the reactor core shell is arranged on the flow channel in a sleeving manner, and the inner cavity of the reactor core shell is communicated with the upper cavity and the lower cavity. And the upper end of the reactor core rod bundle extends out of the upper chamber so as to be connected with a power supply. The reactor core heat exchange test device provided by the embodiment of the invention is high in structural strength and high in test precision, and can reproduce the geometric structure of the pressurized water reactor core to the greatest extent, and the obtained test data is closer to the real condition of the reactor core under the condition of a prototype power station accident.

The invention belongs to the technical field of dispersion fuels, and discloses a method for analyzing a dual non-uniform phenomenon of a plate-shaped fuel element of an RPT in a pressurized water reactor. The RPT method suitable for the fuel plate containing the dispersion particles is optimized and researched. The method comprises the following steps: firstly, evaluating the sensitivity of parameters of a fuel plate containing dispersion particles to the equivalent side length of an RPT model by adopting a quantitative method, determining three important parameters, and analyzing the relationship between the three important parameters and the equivalent side length of the RPT model; finally, a method for rapidly fitting the equivalent side length in the fuel plate containing the dispersion particles is provided, the RPT model can be directly established through a fitting formula, and therefore traditional neutronics calculation software can be rapidly used for calculation to obtain a result. Through numerical verification, for typical plate element dispersion particle fuel problems, the fitting RPT method can achieve high precision, and it is indicated that the method has high application value.

The invention provides a nuclear reactor comprehensive energy system in a data center scene and a control method. The nuclear reactor comprehensive energy system comprises a pressurized water reactor power generation device, an electrolytic cell hydrogen production device, a fuel cell power generation device and an absorption refrigeration device which are communicated with one another, when the power generation of the pressurized water reactor power generation device is sufficient, the pressurized water reactor power generation device is preferentially used for supplying power to the absorption type refrigeration device and supplying surplus power to the electrolytic cell hydrogen production device for hydrogen production; and when the pressurized water reactor power generation device cannot meet the power demand of the absorption type refrigeration device, the fuel cell power generation device supplies power to the absorption type refrigeration device and the electrolytic cell hydrogen production device. The pressurized water reactor is coupled with the wind-light and fuel cell power generation equipment, so that nuclear energy and wind-light resources are fully utilized, and the dual goals of safe and low-carbon efficient power supply are achieved by controlling the full load of the pressurized water reactor and the continuous and safe operation of the electrolytic cell.

This application relates to a fault handling method and apparatus for a nuclear power plant, and relates to the field of nuclear power technology. The method comprises obtaining detection signal sets corresponding to electrical partitions within a pressurized water reactor (PWR), determining target failure modes and temperature correlation values ​​for the PWR based on each detection signal set, and performing fault handling on the PWR based on the target failure modes and temperature correlation values. This method can improve the safety of nuclear power operations.

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 invention discloses a primary circuit washing and foreign matter checking method for a pressurized water reactor nuclear power plant. The primary circuit washing and foreign matter checking method for the pressurized water reactor nuclear power plant comprises the following steps: S10, obtaining the size and morphology of a foreign matter; s20, according to the flowing direction of the operation medium of the loop system, the size of the foreign matter and the intercepting area of the passing equipment, the pipeline where the foreign matter possibly flows through and the position of the equipment where intercepting is possible are obtained; s30, calculating the critical flow velocity of the foreign matter in the primary loop medium; s40, quantitatively judging possible paths and closure points of the foreign matters; and S50, determining a flushing path and a flushing flow according to the possible path and the interception point of the foreign matter. According to the primary loop flushing and foreign matter checking method for the pressurized water reactor nuclear power plant, the foreign matter can be effectively checked, and the checking efficiency can be improved.

The invention discloses a rod position information synchronous monitoring device based on a rod falling test, and relates to the field of pressurized water reactor nuclear power stations, the device comprises a rod control system, a rod position system and interface equipment; the rod control system is used for acquiring an induction current signal output to the corresponding holding coil by each sub-group rod bundle holding case when a rod falling button is pressed down, determining a rod falling reference signal based on the induction current signal, and synchronously transmitting the rod falling reference signal to the rod position system through interface equipment; the rod position system is used for collecting an induced voltage signal of a primary coil in the rod position detector when a rod falling button is pressed down, obtaining rod position measurement information based on the induced voltage signal, and synchronously transmitting the rod position measurement information to the rod control system through interface equipment; and the rod control system and the rod position system synchronously obtain a rod falling test analysis result according to the rod position measurement information and the rod falling reference signal, so that synchronous monitoring of the rod position information is realized. According to the invention, the efficiency of the rod dropping test can be improved and the labor cost is saved.

The invention belongs to the technical field of isotope production, and particularly relates to a 177Lu irradiation production method based on a pressurized water reactor. Comprising the following steps: step 1, packaging Lu2O3 powder in a target box, performing sealing treatment, and storing the target box in a storage tank of a rabbit system outside a reactor core; 2, during the operation period of the pressurized water reactor, transporting the target piece box into the reactor core through a running rabbit system; 3, the target piece box receives neutron irradiation at the axial middle position of the central instrument pipe, and 176Lu reacts to generate 177Lu; 4, when the 177Lu generated by the irradiation of the 176Lu in the target box reaches the designed saturation specific activity, transporting the target box to a post-irradiation target storage tank outside the reactor core; and 5, during the operation period of the pressurized water reactor, repeating the steps 2-4 until the reactor is shut down. The method has the beneficial effects that the single irradiation time of the target boxes can be controlled through analysis, so that a plurality of target boxes can be irradiated in a single refueling period of the pressurized water reactor, and the yield of 177 Lu is greatly improved.

The invention relates to the technical field of pressurized water reactor nuclear power station pipeline hot spot ultrasonic decontamination, in particular to a pressurized water reactor nuclear power station pipeline hot spot ultrasonic decontamination auxiliary device and a using method. According to the scheme, through combination of the electric heater and the ultrasonic vibrator, hot spots of the pipeline are directly heated in the sealed cavity, ultrasonic vibration is generated, and the decontamination effect can be rapidly improved. Large-range energy waste is avoided through local heating, the device is suitable for pipelines of different specifications through the compact modular design, rapid disassembly and assembly are convenient, and the operation difficulty and the time cost are reduced. In addition, the multi-layer sealing design of the sealing ring and the sealing chamber is combined with the application of sealing silica gel, so that the sealing performance under high temperature and high pressure is ensured, and leakage of solution and heat is prevented. The ultrasonic vibrator enhances the chemical decontamination process, especially removal of stubborn corrosives, ensures thorough decontamination, and reduces residual radioactive substances. Due to the advantages, the device has wide application value.

The utility model discloses an indirect heating type fuel rod simulation piece which comprises a cladding tube and an electric heating element, and the cladding tube is a corrosion product deposition test surface; the electric heating element is arranged in the cladding tube and is used for simulating heat generated by a nuclear fission reaction of the fuel pellet; high-purity helium is filled between the cladding tube and the electric heating element, and sealing of the fuel rod simulation piece is achieved through end plug welding at one end of the cladding tube and circumferential welding between the cladding tube and the electric heating element. According to the utility model, the actual structural characteristics and surface material properties of the fuel element are retained to the greatest extent while the surface thermal load state of the pressurized water reactor fuel element is simulated, so that the authenticity of corrosion product deposition test simulation and the reliability of obtaining test data are improved; and the method can also be used for experimental research on scale-caused accelerated corrosion after deposition of surface corrosion products.

The invention discloses a pressurized water reactor nuclear power plant airborne C-14 annual discharge amount determination method, which comprises the following steps: S1, selecting a pressurized water reactor nuclear power plant which is of the same type and has been operated as a reference pressurized water reactor nuclear power plant, and taking C-14 operation discharge monitoring data of the reference pressurized water reactor nuclear power plant as a basis; and S2, calculating the C-14-year discharge amount of the pressurized water reactor nuclear power plant to be determined according to a formula. According to the method, on the basis of C-14 operation emission monitoring data of operated pressurized water reactor nuclear power plants of the same type, the airborne C-14 annual emission amount of the pressurized water reactor nuclear power plants is determined through factors such as reactor power, coolant density, energy utilization rate and radioactive effluent emission management plans; the method for determining the airborne C-14 concentrated emission amount according to the emission management factors of the reference nuclear power plant is clear in principle, simple in process and high in applicability, is suitable for analyzing the expected C-14 annual emission amount of the pressurized water reactor nuclear power plant and is also suitable for determining the conservative C-14 annual emission amount and the emission amount application value.

This invention relates to the monitoring and diagnosing of nuclear power plants for its thermal performance using the NCV Method. Its applicability comprises any nuclear reactor such as used for research producing a useful output. Its greatest applicability lies with conventional Pressurized Water Reactor and Boiling Water Reactor nuclear plants generating an electric power. Its teachings of treating fission as an inertial process, a phenomena which is self-contained following incident neutron capture, allows the determination of an absolute neutron flux. This process is best treated by Second Law principles producing a total fission exergy. This invention also applies to the design of fusion thermal systems regards the determination of its Second Law viability and absolute plasma flux.

The invention relates to an on-line setting method and system for a spring loading valve of a pressurized water reactor nuclear power plant, which can be used for finishing the adjustment of a spring preloading pressure value under the condition that the valve does not need to be disassembled, so that the time required for maintenance is greatly reduced. And secondly, compared with a traditional method for replacing the whole valve, the method avoids high overall purchase cost of the valve and complex installation engineering, and the maintenance cost is effectively saved. Meanwhile, manpower and material resource investment is reduced, and the overall maintenance cost is further reduced. Furthermore, due to the fact that complex operation such as cutting of connecting pipelines is not needed, potential safety hazards possibly occurring in the maintenance process are remarkably reduced, and therefore the safety of maintenance personnel and stable operation of the system are guaranteed.