CN103985422A - Active and passive nuclear steam supplying system based on 177 reactor core and nuclear power station thereof - Google Patents
Active and passive nuclear steam supplying system based on 177 reactor core and nuclear power station thereof Download PDFInfo
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- CN103985422A CN103985422A CN201410105700.7A CN201410105700A CN103985422A CN 103985422 A CN103985422 A CN 103985422A CN 201410105700 A CN201410105700 A CN 201410105700A CN 103985422 A CN103985422 A CN 103985422A
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
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- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
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Abstract
The invention relates to an active and passive nuclear steam supplying system based on 177 reactor core. The nuclear steam supplying system comprises a nuclear reactor core and a reactor coolant system, and is characterized in that the nuclear reactor core comprises 177 nuclear fuel assemblies with the active length of 12-14 feet; and the reactor coolant system comprises a reactor pressure vessel, a main pipe connecting an inlet and an outlet of a reactor coolant, a main pump, a steam generator, a pressure stabilizer and a pressure relief tank. The invention also relates to a nuclear power station employing the above active and passive nuclear steam supplying system based on the 177 reactor core. The power of the assembling unit of the nuclear power station is 1000-1400 MWe, the average availability is more than or equal to 90%, the maximum ground acceleration is 0.3 g, and a containment has a double-layer steel-made structure for resisting bump of large business airplanes. The active and passive nuclear steam supplying system based on the 177 reactor core has the functions of alleviating and preventing severe accidents, and a core measuring meter penetrates into the reactor pressure vessel from top to bottom, so that the active and passive nuclear steam supplying system possesses an active manner and passive manner combined residual heat removal system and a digitalizer control diversity protection system.
Description
Technical field
The present invention relates to three generations's nuclear power technology field, be specifically related to a kind of based on actively adding of 177 reactor cores non-active nuclear steam supply system and nuclear power station thereof.
Background technology
Nuclear reactor design is one of key Design content of nuclear power station.Fuel assembly is the important component part of nuclear reactor.The main task of nuclear reactor design is to provide from the angle of nuclear reactor physics the nuclear reactor that meets pressurized-water reactor nuclear power plant general design requirement, comprises and determines that fuel assembly number, fuel assembly are in the layout of nuclear reactor etc.Current million kilowatt three loop pressurized-water reactor nuclear power plant reactor cores consist of 157 fuel assemblies, and its core power density is larger, and thermal technology's margin of safety is relatively low.
Nuclear power station, under major accident operating mode, lacked prevention and the mitigation strategy of major accident in the past.Lack high-order exhaust, cannot be under accident conditions, discharge the non-condensable gas that reactor pressure vessel top is gathered, thereby the impact that causes these not concretive gas to conduct heat on reactor core, cannot guarantee to only have unique carbonated drink interface in reactor coolant loop, cause bad damage sequence; Lack a loop fast pressure relief, cannot under major accident, carry out rapid pressure relief function, thereby can not reduce the risk that under major accident, high-voltage fuse heap brings, so that the high-pressure smelting thing jet phenomenon that occurs threatening containment integrity; Disappearance steam generator secondary side passive residual heat removal function, cannot lose under the accident conditions of active residual heat of nuclear core discharge ability at generation station blackout and power plant, so that can not derive for a long time residual heat of nuclear core, affects the safety of reactor.Also lack after LOCA the function such as stoppage in transit main pump and operator's nonintervention in 30 minutes automatically simultaneously, can not guarantee better the safe operation of nuclear power station.
In addition, in the nuclear power station that existing two generations add, for monitoring the In-core Instrumentation instrument of reactor operation correlation parameter, generally from reactor core bottom, insert, need to be in pressure vessel bottom opening, along with improving constantly that nuclear reactor safety is required, In-core Instrumentation instrument need to be inserted from reactor core top, to cancel pressure vessel bottom opening, and then the security that improves reactor.
Meanwhile, for realizing Security Target, current nuclear power station is provided with residual heat of nuclear core and discharges system, and implementation has two kinds: the enabling fashion of traditional M310 type, and the western room AP1000 non-enabling fashion that is representative.Non-ly actively refer to that equipment or system only rely on the modes relevant to spontaneous phenomenon such as gravity, density, Natural Circulation and drive, and without introducing other propulsion system, can greatly reduce the equipment failure probability causing because of power machine fault, improve the reliability of security system.Active equipment has the advantages such as power is strong, pressure is high, flow is large, compact conformation.Along with the development of nuclear power technology is with strict, safety coefficient after reactor operation, shutdown and under accident conditions is more and more higher, therefore need the residual heat removal system of a kind of combination enabling fashion and non-enabling fashion advantage, thereby improve the safety and reliability of nuclear power station.
At present, nuclear power station adopts digitizing instrument control system (DCS) as the control system of master-control room one after another both at home and abroad, and its reliability and security become the problem that final user is concerned about the most.Nuclear power plant's digitizing instrument control system (DCS), is generally divided into two platforms, the main protection system platform of safe level and non-security level system platform.Reactor main protection system (RPS: comprise reactor emergency shut-down function, engineered safeguards features function) adopts digitizing instrument control; with respect to traditional simulation instrument control system; numerical value collection is more accurate, and man-machine interface is more friendly, is conducive to improve efficiency, the safety and reliability of nuclear power plant's operation.But based on digitized reactor main protection system, may there is software common mode failure (Software Common Cause Failure, SWCCF), thereby cause main protection system complete failure.If there is software common mode failure, and the superpose imaginary anticipated transient of generation or design basis accident (hereinafter to be referred as: software common mode stack accident), accident cannot be alleviated, and will cause serious consequence. simultaneously
Summary of the invention
The technical matters that will solve of the present invention is to provide a kind of employing 177 reactor core arrangements; have and alleviate and prevention major accident function; In-core Instrumentation instrument penetrates reactor pressure vessel from top to bottom, has and combines the active remaining residual heat removal system of non-enabling fashion and nuclear steam supply system and the nuclear power station thereof of digitizing instrument control Protection of Diversity system.
In order to solve the problems of the technologies described above, technical scheme of the present invention is, a kind of based on non-active nuclear steam supply system of actively adding of 177 reactor cores, comprises nuclear reactor, reactor coolant loop, described nuclear reactor comprises that 177 active lengths are the nuclear fuel assembly of 12 to 14 feet;
Described reactor coolant loop comprises reactor pressure vessel, main pump, steam generator, voltage stabilizer, release case and the coolant entrance of coupled reaction core pressure vessel and the main pipeline of outlet; Described main pipeline comprises cold section, hot arc and transition section, described hot arc is connected to primary side of steam generator entrance, described primary side of steam generator outlet is connected with transition section pipeline one end, and the other end of described transition section is communicated with cold section, thereby forms reactor-loop; Described voltage stabilizer lower end is connected on hot arc by fluctuation pipe; On described transition section and fluctuation pipe, be provided with LBB leakage detector.
Described reactor coolant loop also comprises main pump, release case;
Described main pump is arranged on described cold section; Between the entrance and exit of described main pump, be provided with pressure tester; This voltage stabilizer top is provided with the first safe ozzle and the second safe ozzle, and described the first safe ozzle is connected with release main pipeline one end by the first pressure relief pipeline; The described release main line other end is communicated with release case portion; Described the second safe ozzle is connected on release main line by the second pressure relief pipeline; On described the second pressure relief pipeline, be provided with pressurizer safety valve; Described the first pressure relief pipeline and release main pipeline junction are provided with fast pressure relief valve, are provided with the valve of controlling flow on the first pressure relief pipeline between this fast decompression valve and described the first safe ozzle; Described reactor pressure vessel top is connected on described release main line by pressure vessel top gas exhaust piping; The vent valve of some series connection is set on the gas exhaust piping of described pressure vessel top.
Described reactor pressure vessel volume is 50～80m
3.
Described 177 fuel assemblies are arranged to the reactor core of 15 row, 15 row; Each nuclear fuel assembly is square fuel assembly, and it comprises 264 nuclear fuel rods, 24 control rod guide pipes and the 1 measurement instrument pipe that is arranged in 17 row, l7 row.
Described nuclear reactor thermal technology allowance is greater than 15%; The first circulation refulling cycle is 18～24 months.
In the first circulation of reactor core, described 177 nuclear fuel assemblies comprise several burnable poison fuel assemblies, described burnable poison fuel assembly comprises some burnable poison fuel rods, described burnable poison fuel rod comprises several burnable poison fuel pellets, and described burnable poison fuel pellet is by Gd
2o
3with UO
2mix and sintering formation.
Gd in described burnable poison fuel pellet
2o
3percentage by weight be 2%～10%.
Described reactor pressure vessel, steam generator, main pump, voltage stabilizer, main pipeline designed life is 60 years.
Described reactor pressure vessel top is also provided with missile roof shielding steel plate and lifts the integrated heap top system of enclosing cylinder.
Also comprise that actively adding non-active heat discharges system, this actively adds non-active heat discharge system and comprises active system and passive system, and described active system comprises safety injection system, active Reactor cavity flooding system, safety shower system, steam generator auxiliary feedwater system; Described passive system comprises high-order safety injection system, non-active Reactor cavity flooding system, passive containment thermal conduction system, Passive residual heat removal system; Described safety injection system, active Reactor cavity flooding system, safety shower system, high-order safety injection system, non-active Reactor cavity flooding system are all communicated with reactor-loop by pipeline; Described steam generator auxiliary feedwater system, Passive residual heat removal system are all communicated with reactor secondary circuit by pipeline; Passive containment thermal conduction system is arranged on containment top.
Described safety injection system comprises the first water source and the first active pipeline, and described first active pipeline one end is connected the other end with the first water source and is connected on reactor-loop.
Described active Reactor cavity flooding system comprise second water source, one end be connected with second water source the other end be connected to reactor cavity bottom, one end connects the outside fire water supply other end and is connected to the fire extinguisher canvas hose on the second active pipeline.
Described safety shower system comprises that the 3rd water source, the circular spray tube that is arranged on containment inner top, one end are connected with the 3rd water source the spray pipeline that the other end is connected with this circular spray tube.
Described steam generator auxiliary feedwater system comprises that the 4th water source, one end are connected the 3rd active pipeline on the main feed water pipe road that the other end is connected to reactor secondary circuit with the 4th water source.
Described high-order safety injection system comprises that peace note case, one end of being arranged on higher than reactor-loop position are connected with peace note case bottom the first non-active pipeline that the other end is connected with cold section of reactor-loop.
Described non-active Reactor cavity flooding system comprises that non-active Reactor cavity flooding case, one end of being arranged on higher than reactor core position are connected the second non-active pipeline that the other end is connected to reactor cavity bottom with non-active Reactor cavity flooding case.
Described passive containment thermal conduction system comprises the first heat-exchanging water tank, is arranged on the second heat interchanger in containment, and described the second heat interchanger is communicated with the first heat-exchanging water tank by pipeline.
Described Passive residual heat removal system comprises and is arranged on the second heat-exchanging water tank, is arranged on the 3rd heat interchanger in this second heat-exchanging water tank, and the water inlet of described the 3rd heat interchanger and water delivering orifice are connected respectively on the main feed water pipe road and main steam line of reactor secondary circuit.
Described the first water source, second water source, the 3rd water source are and are arranged on built-in material-changing water tank in containment vessel; Described the 4th water source is the auxiliary feed-water tank being arranged on outside containment vessel.
Also comprise instrument control system; Described instrument control system comprises reactor core measuring system, Protection of Diversity system DAS; Described reactor core measuring system comprises that several insert the detector assembly of reactor core from pressure vessel top, and described detector assembly comprises level sensor assembly and neutron-hygrosensor assembly; The signal output part of described level sensor assembly and neutron-hygrosensor assembly passes respectively containment and is connected to Core cooling monitor signal treatment facility and reactor core neutron flux signal handling equipment.
Described level sensor assembly comprises inner two sensitive element M1 for hollow structure and M2; The inside of described M1 is provided with two K type thermopair HT1 and UHT1, and for the electric heater HE1 to HT1 heating; The inside of described M2 is provided with two K type thermopair HT2 and UHT2, and for the electric heater HE2 to HT2 heating; Described HT1 and HT2 are drive end, the reference edge that UHT2 is HT1, the reference edge that UHT1 is HT2; The inside of described sensitive element Ml and M2 is all filled with insulation course; The material of described insulation course is magnesium oxide.
Described neutron-hygrosensor assembly comprises package shell, self-power neutron detector group, internal seal structure, connector, end part seal plug; Described self-power neutron detector group is fixed in described package shell, and it comprises N self-power neutron detector, a described N self-power neutron detector along described package shell axially from bottom to top evenly interval arrange; The vertical range of the lower surface, sensitive section of centre distance reactor fuel assemblies active region of i rhodium self-power neutron detector is from bottom to top
i=1 wherein, 2,3 ..., N; H is the axial height of reactor fuel assemblies active region; Described connector sealing is fixed on the upper end of described package shell; Described end part seal plug sealing is fixed on the lower end of described package shell; Described internal seal structure sealing is fixed on the inside of the described package shell that is positioned at described self-power neutron detector group top; The sheathed cable of described N rhodium self-power neutron detector is tied up and fixes, and axially extends upward along described package shell, through described internal seal structure, is finally fixed on the stitch of described connector; Also comprise and be fixed on the Pt100 four-wire system thermometer that is positioned at described internal seal structure upper area in described package shell, described Pt100 four-wire system thermometer comprises 4 lead-in wires; Described 4 lead-in wires are connected respectively on 4 stitch of described connector; Also comprise and be fixed on the thermopair that is positioned at described internal seal structure lower zone in described package shell; Described thermopair is connected on the stitch of described connector by thermopair extended line.
Described Protection of Diversity system DAS adopts the equipment that is different from reactor protection system, to prevent the generation of the common mode failure identical with reactor protection system.
A nuclear power station, adopt described in claim 1 to 22 any one based on non-active nuclear steam supply system of actively adding of 177 reactor cores;
Its power of the assembling unit 1000～1400MWe, average availability is more than or equal to 90%, and maximum ground acceleration is 0.3g, and containment is that double-deck steel design is clashed into Chinese People's Anti-Japanese Military and Political College's type business aircraft.
Its nuclear reactor probability of damage CDF is lower than 10
-6/ heap year, early stage a large amount of radiomaterials discharge probability LERF lower than 10
-7/ heap year.
Beneficial effect of the present invention:
(1) reactor core thermal technology safety allowance is greater than 15%;
(2) can effectively prevent and alleviate major accident, comprising stoppage in transit main pump and operator's nonintervention in 30 minutes automatically after the exhaust of the high point of voltage stabilizer fast pressure relief, reactor pressure vessel, LOCA; By LBB leakage monitoring, the ruuning situation of effective monitoring main pipeline and fluctuation pipe, gives corresponding safety assessment, thereby improves reactor operation reliability;
(3) In-core Instrumentation instrument need to be inserted from reactor core top, to cancel pressure vessel bottom opening, and then the security that improves reactor;
(4) combine active and non-enabling fashion and carry out residual heat of nuclear core discharge, improve the safety and reliability of nuclear power station;
(5) digitizing instrument control Protection of Diversity system adopts the equipment that is different from reactor protection system, to prevent the generation of the common mode failure identical with reactor protection system, determines the Protection of Diversity of reactor.
Accompanying drawing explanation
Fig. 1 the present invention is based on reactor coolant loop schematic diagram in non-active nuclear steam supply system of actively adding of 177 reactor cores;
Fig. 2 the present invention is based in non-active nuclear steam supply system of actively adding of 177 reactor cores, actively to add non-active heat-extraction system schematic diagram;
Fig. 3 the present invention is based on reactor core measuring system schematic diagram in non-active nuclear steam supply system of actively adding of 177 reactor cores;
In figure: 001-LBB leakage detector, 002-voltage stabilizer, 003-main pump, 004-reactor pressure vessel, 005-release case, 006-level sensor assembly, 007-neutron-hygrosensor assembly, 008-reactor core neutron flux signal handling equipment, 009-Core cooling monitor signal treatment facility, 1-reactor core, 2-reactor-loop, cold section of 201-, 202-hot arc, 203-transition section, 3-steam generator, 4-reactor secondary circuit, 401-main feed water pipe road, 402-main steam line, the active pipeline of 5-first, 501-the first pump, 6-reactor cavity, the active pipeline of 7-second, 701-the second pump, 702-fire extinguisher canvas hose, 8-spray pipeline, 801-circular spray tube, 802-the 3rd pump, First Heat Exchanger 803, the active pipeline of 9-the 3rd, 901-pneumatic pump, 902-electrodynamic pump, 10-containment, the built-in material-changing water tank of 11-, 12-auxiliary feed-water tank, 13-peace note case, the non-active pipeline of 14-first, the non-active Reactor cavity flooding case of 15-, the non-active pipeline of 16-second, 17-the first heat-exchanging water tank, 18-the second heat interchanger, 19-the first inlet pipeline, 20-the first outlet conduit, 21-steam-water separator, 22-the second heat-exchanging water tank, 23-the 3rd heat interchanger, 24-the second inlet pipeline, 25-the second outlet conduit.
Embodiment
Below in conjunction with drawings and Examples, the present invention is described further.
The present invention is a kind of based on non-active nuclear steam supply system of actively adding of 177 reactor cores, comprises nuclear reactor, reactor coolant loop, actively adds non-active heat and discharge system, instrument control system;
Described nuclear reactor comprises that 177 active lengths are the nuclear fuel assembly of 12 to 14 feet; In forming 177 nuclear fuel assemblies of this reactor core, by 157 nuclear fuel assemblies, be arranged in parallel with each other and form the reactor core of 15 row, the 15 existing 1,001,000 watts of level three loop presurized water reactors that are listed as, the surrounding that separately has the existing pressurized water reactor core fuel assembly of being arranged in parallel within of 20 nuclear fuel assembly symmetries periphery, has formed 15 row that are comprised of 177 nuclear fuel assemblies, the new reactor core of 15 row.Wherein, the 1st row and the 15th row of reactor core respectively have 5 of nuclear fuel assemblies, the the 2nd and the 14th row respectively has 9 of nuclear fuel assemblies, the the 3rd and the 13rd row respectively has 11 of nuclear fuel assemblies, the the 4th and the 12nd row respectively has 13 of nuclear fuel assemblies, the the 5th and the 11st row respectively has 13 of nuclear fuel assemblies, and the 6th to the 10th respectively has 15 of nuclear fuel assemblies; Each nuclear fuel assembly is square fuel assembly, and it comprises 264 nuclear fuel rods, 24 control rod guide pipes and the 1 measurement instrument pipe that is arranged in 17 row, l7 row;
Described nuclear reactor is compared with active service reactor core, and in the situation that nuclear fuel rod active length is identical, reactor core has reduced by 11.3% power density than active service reactor core, and reactor core thermal technology allowance is greater than 15%; The first circulation refulling cycle is 18～24 months;
Be specially, in the first circulation of reactor core forming at 177 fuel assemblies, use Gd
2o
3as burnable poison, improve fuel assembly initial enrichment, fuel arranged assembly, in the position of heap in-core, makes the length of the cycle of first circulation reach 18～24 months length of reloading; Gd
2o
3as solid combustible poisonous substance, be structurally and UO
2mixing sintering are at UO
2in pellet, therefore use Gd
2o
3as solid combustible poisonous substance, do not take the position of guide pipe in fuel assembly, control rod is to insert guide pipe position, therefore uses Gd
2o
3as solid combustible poisonous substance, do not interfere with the Position Design of control rod, compare the solid combustible poisonous substance of traditional use borosilicate glass and the situation that control rod Position Design interferes, the dirigibility that has improved to the full extent control rod Position Design; Due to Gd
2o
3be and UO
2pellet is sintering together, do not increase the refuse amount outside spentnuclear fuel, and borosilicate glass is as inserting type assembly, need to carry out extra packing and processing, can additionally cause radioactive waste.Meanwhile, Gd
2o
3manufacturing process comparatively simple easily, and borosilicate glass burnable poison is as inserting type assembly, comparatively complicated on manufacturing.In first circulation, contain Gd in gadolinia fuel pellet
2o
3percentage by weight can be from 2%～10%.
Described reactor coolant loop, as shown in Figure 1, comprises reactor pressure vessel 004, the main pipeline of coupled reaction reactor coolant entrance and exit, main pump 003, steam generator 3, voltage stabilizer 002, release case;
Described reactor pressure vessel 004 volume is 50～80m
3;
Described main pipeline comprises cold section 201, hot arc 202 and transition section 203, described main pump 003 is arranged on described cold section 201, described hot arc 202 is connected to steam generator 3 primary side entrances, and described steam generator 3 primary side outlets are communicated with main pump 003 by transition section pipeline 203; Thereby form a loop of nuclear reactor; Between the entrance and exit of described main pump 003, be provided with pressure tester;
Described voltage stabilizer 002 lower end is connected on hot arc 202 by fluctuation pipe; This voltage stabilizer 002 top is provided with the first safe ozzle and the second safe ozzle, and described the first safe ozzle is connected with release main pipeline one end by the first pressure relief pipeline; The described release main line other end is communicated with release case 005 top; Described the second safe ozzle is connected on release main line by the second pressure relief pipeline; On described the second pressure relief pipeline, be provided with pressurizer safety valve;
On described transition section and fluctuation pipe, be provided with LBB leakage detector 001;
Described the first pressure relief pipeline and release main pipeline junction are provided with fast pressure relief valve, are provided with the valve of controlling flow on the first pressure relief pipeline between this fast decompression valve and described the first safe ozzle; Described the first pressure relief pipeline can arrange redundancy branch road, and many described the first pressure relief pipelines in parallel are set;
Described reactor pressure vessel 004 top is connected on described release main line by pressure vessel top gas exhaust piping; Described pressure vessel top gas exhaust piping comprises some branch roads in parallel, and the vent valve of some series connection is set on every branch road;
Described reactor coolant loop can be realized after the 004 high some exhaust of voltage stabilizer 002 fast pressure relief, reactor pressure vessel, LOCA the function of stoppage in transit main pump 003 and operator's nonintervention in 30 minutes automatically.
At described reactor coolant loop, reactor pressure vessel 004, steam generator 3, main pump 003, voltage stabilizer 002, main pipeline designed life is 60 years;
Described reactor pressure vessel 004 top is also provided with missile roof shielding steel plate and lifts the integrated heap top system of enclosing cylinder, effectively reduces and reloads the time, improves economy;
Reactor coolant loop, reactor pressure vessel 004 neutron moisture buret end socket disposed thereon, the security that improves reactor pressure vessel 004 structure, the probability of Lower head failure under reduction accident conditions; Main pipeline and Pressurizer surge line adopt LBB technology, and special leakage monitoring system is set, and cancel anti-rejection limiter; Fluctuation pipe is arranged and can effectively be alleviated thermal stratification; Cancel thermometric bypass subsystem, simplification system arranges; Employing volume is 50～80m
3voltage stabilizer, improve steady pressure of system ability.
Describedly actively add non-active heat and discharge system, as shown in Figure 2, comprise the reactor core 1 that is arranged in reactor pressure vessel 004, the reactor-loop 2, the steam generator 3 that by cold section 201, hot arc 202 and reactor core 1 inner flow passage, are formed and the reactor secondary circuit 4 being formed by steam generator main feed water pipe road 401 and main steam line 402, also comprise active system and passive system, described active system comprises safety injection system, active Reactor cavity flooding system, safety shower system, steam generator auxiliary feedwater system; Described passive system comprises high-order safety injection system, non-active Reactor cavity flooding system, passive containment thermal conduction system, Passive residual heat removal system;
Described safety injection system, active Reactor cavity flooding system, safety shower system, high-order safety injection system, non-active Reactor cavity flooding system are all connected with reactor-loop 2 by pipeline; Described steam generator auxiliary feedwater system, Passive residual heat removal system are all connected with reactor secondary circuit 4 by pipeline; Passive containment thermal conduction system is arranged on containment 10 described in containment 10 tops and adopts double-deck steel design;
Described safety injection system comprises that the first water source, one end are connected with the first water source that the other end is connected to cold section 201 of reactor-loop or/and the first active pipeline 5 of hot arc 202 also comprises the first pump 501 being arranged on this first active pipeline 5.What this safety injection system generally adopted is cold section of 201 injection systems, takes cold section 202 injection mode simultaneously after entering long-term cooling stage.
Described active Reactor cavity flooding system comprises that second water source, one end are connected the second active pipeline 7 that the other end is connected to reactor cavity 6 bottoms with second water source, also comprises that the second pump 701, one end of being arranged on this second active pipeline connect the outside fire water supply other end and be connected to the fire extinguisher canvas hose 702 on the second active pipeline 7 between described second water source and the second pump 701 and be arranged on the solenoid valve on affiliated fire extinguisher canvas hose 702.
Described safety shower system comprises the 3rd water source, is arranged on the circular spray tube 801 of containment 10 inner tops, one end is connected the other end and the spray pipeline 8 that this circular spray tube 801 is connected with the 3rd water source, also comprises that series connection is arranged on the 3rd pump 802 and the First Heat Exchanger 803 on this spray pipeline successively.
Described steam generator auxiliary feedwater system comprises that the 4th water source, one end are connected the 3rd active pipeline 9 on the main feed water pipe road 402 that the other end is connected to reactor secondary circuit 4 with the 4th water source, and also comprising is connected in parallel to each other is arranged on many groups pneumatic pump 901 and the electrodynamic pump 902 on the 3rd active pipeline 9 described in this.Here adopt the allocation plan of 2 50% electrodynamic pump+2,50% pneumatic pump, thereby improved the reliability of auxiliary feedwater, and can meet single failure criteria.
Above-mentioned the first water source, second water source, the 3rd water source are the built-in material-changing water tank 11 being arranged in containment vessel 10; Described the 4th water source is the auxiliary feed-water tank 12 being arranged on outside containment vessel 10.
Described high-order safety injection system comprises that peace note case 13, one end of being arranged on higher than reactor-loop 2 positions are connected with peace note case 13 bottoms the first non-active pipeline 14 that the other end is connected with cold section 201 of reactor-loop 2 and/or hot arc 202, also comprise that series connection is successively set on solenoid valve and the non-return valve on the first non-active pipeline 14 from top to bottom.Here can establish a plurality of peace note casees 13, as 3, and logical super cooled sect 201 piii reactor reactor cores 1.
Described non-active Reactor cavity flooding system comprises that non-active Reactor cavity flooding case 15, one end of being arranged on higher than reactor core 1 position are connected the second non-active pipeline 16 that the other end is connected to reactor cavity 6 bottoms with non-active Reactor cavity flooding case 15; Described the second non-active pipeline 16 is the pipeline of connecting from top to bottom after two groups of first parallel connections, one group of top parallel pipeline one end is connected on the different vertical height and position of described non-active Reactor cavity flooding case 15, and the other end collects and is a pipe and connects with one group of pipeline on the lower; Solenoid valve and non-return valve that on every pipeline of one group of parallel pipeline on the lower, series connection arranges successively from top to bottom, and finally collect and be a pipe and be connected to reactor cavity 6 bottoms.
Described passive containment thermal conduction system comprises that containment 10 outsides are set to be stored the first heat-exchanging water tank 17 of cold water, be arranged on the second heat interchanger 18 in containment 10, connect the first inlet pipeline 19 and first outlet conduit 20 of the first heat-exchanging water tank 17 and the second heat interchanger 18; First inlet pipeline one end is connected to the first heat-exchanging water tank 17 bottoms, and the other end is connected to the import of heat exchanger tube in the second heat interchanger 18; First outlet conduit 20 one end are connected with the steam-water separator 21 in being arranged on the first heat-exchanging water tank 17, and the other end is connected to the outlet of heat exchanger tube in the second heat interchanger 18.
Described Passive residual heat removal system comprises being arranged on and containment 10 outsides are set store the second heat-exchanging water tank 22 of cold water, is arranged on the 3rd heat interchanger 23 in this second heat-exchanging water tank 22, connects the second inlet pipeline 24 and second outlet conduit 25 of the second heat-exchanging water tank 22 and the 3rd heat interchanger 23; Described second inlet pipeline 24 one end are connected with the import of heat exchanger tube in the 3rd heat interchanger 23, and the other end is connected on reactor secondary circuit 4 main steam lines 401; Described second outlet conduit 25 one end are connected with the outlet of heat exchanger tube in the 3rd heat interchanger 23, and the other end is connected on reactor secondary circuit 4 main feed water pipe roads 402; And, on this second outlet conduit 25, be disposed with solenoid valve and non-return valve from top to bottom.Above-mentioned the first heat-exchanging water tank 17 and the second heat-exchanging water tank 22 are positioned at the peripheral upside of containment 10, and they can be designed to be a monoblock type water tank, and this monoblock type water tank can be designed to the annular water tank structure around containment 10, has simplified equipment.
Above-mentioned the first pump 501, the second pump 701 and the 3rd pump 802 is the equipment of the ripe application of existing power plant, by can start action under relevant trigger pip, first active pipeline the 5, second active pipeline 7 and spray pipeline 8 are all fetched water and realize Safety Injection, Reactor cavity flooding and spray function from the built-in material-changing water tank 11 of containment 10.Different from existing design, built-in material-changing water tank 11 is built as an one-piece construction in containment 10.The peace note case 13 of high-order safety injection system is in reactor-loop 2 pressure automatic opens check valve when low, by water piii reactor reactor core 1.
Above-mentioned steam generator auxiliary feedwater system is under accident conditions, and when main water work can not be worked, this system supplies water to steam generator 3, and to derive the waste heat in reactor, the steam of generation enters atmosphere.
Above-mentioned active Reactor cavity flooding system, after the heap reactor core 1 damage accident that reacts, is first fetched water by built-in material-changing water tank 11 or fire water, and actively piii reactor heap chamber 6, realizes lasting cooling.When active Reactor cavity flooding system is unavailable because station blackout or emergency diesel-oil machine lost efficacy, rely on gravity by the water piii reactor heap chamber 6 in the non-active Reactor cavity flooding case 15 in non-active Reactor cavity flooding system, realize the injection of chilled water.This system can prevent reactor core 1 fused mass burn through pressure vessel, guarantees that second physical barrier is to a large amount of radioactive containment roles.
Above-mentioned Passive residual heat removal system, under accident conditions, need to carry out while utilizing steam generator 3 to discharge the function of reactor core 1 waste heat, reactor-loop 2 hot water pass through steam generator heat-transfer pipe, heat is transmitted to reactor secondary circuit 4, make the feedwater boiling of reactor secondary circuit 4 become steam, because main steam line under accident conditions 402 is closed, steam is subject to the effect of steam generator 3 interior elevated pressures, along pipeline, enter the 3rd heat interchanger 23 in Heat Discharging System of Chinese, be immersed in the 3rd Tube Sheet of Heat Exchanger 23 heat exchanger tubes in the second heat-exchanging water tank 22 is vapour, pipe is outer is water, steam-condensation, condensate water flows out the 3rd heat interchanger 23 heat exchanger tubes under Action of Gravity Field, along heat exchange pipeline, flow back to the main feed water pipe road 401 of steam generator 3, reenter steam generator 3, maintain the water level in steam generator 3.Pattern completion circulation thus, the refrigerating function of 4 pairs of reactor-loops 2 of realization response heap secondary circuit, makes the smooth decrease temperature and pressure of reactor-loop 2, thereby finally makes nuclear power station enter the safe condition of cold shut.
Above-mentioned passive containment thermal conduction system, utilization is built in the second heat interchanger 18 in containment 10, convection current between condensation by water vapour on the second heat interchanger 18, mixed gas and this second heat interchanger 18 and radiation heat transfer are realized cooling in containment 10, by flowing of water in heat exchanger tube in the second heat interchanger 18, continuously the heat in containment 10 is taken to outside containment 10, utilize density difference that the temperature difference of water in the first heat-exchanging water tank 17 outside containment 10 causes to realize non-passive safety shell heat and discharge.
Built-in material-changing water tank 11 of the present invention is positioned at containment, has reduced the impact of outside disaster on material-changing water tank security, has improved the reliability at emergent water source after accident, has improved power plant safety.Have an accident in situation, if adopt external mode, peace note and spray system need to carry out blocked operation under liquid level gauge coordinates, and interior postpone does not need the switching in peace note, peace water spray source.Because built-in material-changing water tank 11 is using the unique active Safety Injection after accident, safety shower source, operation after can minimizing accident, avoided contingent mistake, reduced the potential risk of system running pattern switching failure, thereby improved the reliability of system, strengthened the security of power plant.
This built-in material-changing water tank 11 is positioned at lowest part, conveniently collects the water source bringing from container spray, pipeline cut, after collecting in the lump, via pump, to safety injection system, safety shower system, active Reactor cavity flooding system, supplies with cooling water.Built-in material-changing water tank 11 as the source, water source of above three systems, has played the effect of simplified apparatus uniformly.
In conventional security injected system, fill pump and double as high-pressure safety injection pump simultaneously, the present invention is separated with peace pouring functions by filling in conventional security injected system, cancels high-pressure safety injection pump, presses safety injection pump, i.e. described the first pump 501 in employing.Traditional safety injection system is when peace note signal occurs, pump is switched to peace injection-molded from upper mold filling formula, and this handoff procedure need operate a large amount of valves, will have influence on the reliability of system.After special-purpose middle pressure safety injection pump is set, carry out function singleness, can improve the reliability of system.
In this, press safety injection pump to reduce injection pressure head (becoming pressure from high pressure), error starting accident in the time of can effectively preventing high-pressure safety injection, avoid reactor-loop 2 hypertonia, steam generator 3 spill-overs that also can alleviate or avoid reactor-loop 2 hypertonia under steam generator 3 heat-transfer pipe break accidents and may cause, thus radiomaterial reduced under this accident to the possibility of environmental emission;
For design basis accident, beyond design basis accident, or even major accident, only depend on the operation of passive system, can in 72 hours, not need to rely on operator's manual intervention and realize reactor core protection requirement.
Described instrument control system comprises reactor core measuring system, Protection of Diversity system DAS, core instrument system, measurement system, reactor protection system, reactor control system, rod control and rod position system, excellent power-supply system, reactor loosening part and vibration monitor system;
As shown in Figure 3, described reactor core measuring system comprises that several insert the detector assembly of reactor core from pressure vessel top, and described detector assembly comprises level sensor assembly 006 and neutron-hygrosensor assembly 007; The signal output part of described level sensor assembly 006 and neutron-hygrosensor assembly 007 passes respectively containment and is connected to Core cooling monitor signal treatment facility 009 and reactor core neutron flux signal handling equipment 008;
Described level sensor assembly comprises inner two sensitive element M1 for hollow structure and M2; The inside of described M1 is provided with two K type thermopair HT1 and UHT1, and for the electric heater HE1 to HT1 heating; The inside of described M2 is provided with two K type thermopair HT2 and UHT2, and for the electric heater HE2 to HT2 heating; Described HT1 and HT2 are drive end, the reference edge that UHT2 is HT1, the reference edge that UHT1 is HT2; The inside of described sensitive element Ml and M2 is all filled with insulation course; The material of described insulation course is magnesium oxide.
Described neutron-hygrosensor assembly comprises package shell, self-power neutron detector group, internal seal structure, connector, end part seal plug; Described self-power neutron detector group is fixed in described package shell, and it comprises N self-power neutron detector, a described N self-power neutron detector along described package shell axially from bottom to top evenly interval arrange; The vertical range of the lower surface, sensitive section of centre distance reactor fuel assemblies active region of i rhodium self-power neutron detector is from bottom to top
i=1 wherein, 2,3 ..., N; H is the axial height of reactor fuel assemblies active region; Described connector sealing is fixed on the upper end of described package shell; Described end part seal plug sealing is fixed on the lower end of described package shell; Described internal seal structure sealing is fixed on the inside of the described package shell that is positioned at described self-power neutron detector group top; The sheathed cable of described N rhodium self-power neutron detector is tied up and fixes, and axially extends upward along described package shell, through described internal seal structure, is finally fixed on the stitch of described connector.Also comprise and be fixed on the Pt100 four-wire system thermometer that is positioned at described internal seal structure upper area in described package shell, described Pt100 four-wire system thermometer comprises 4 lead-in wires; Described 4 lead-in wires are connected respectively on 4 stitch of described connector.Also comprise and be fixed on the thermopair that is positioned at described internal seal structure lower zone in described package shell; Described thermopair is connected on the stitch of described connector by thermopair extended line.
Adopt self-supporting moderate energy neutron flux-sensing element to realize the continuous monitoring of reactor core neutron flux, adopt hot conduction-type liquid level sensor to realize the monitoring of reactor core key point liquid level, there is powerful and on-line monitoring computing power.
Described Protection of Diversity system DAS adopts the equipment that is different from reactor protection system, to prevent the generation of the common mode failure identical with reactor protection system.
Adopt an above-mentioned nuclear power station that actively adds non-active nuclear steam supply system, its power of the assembling unit 1000～1400MWe, average availability is more than or equal to 90%, and design maximum ground acceleration 0.3g, arranges double containment, and Chinese People's Anti-Japanese Military and Political College's type business aircraft clashes into.
Owing to being provided with, actively add non-active nuclear steam supply system, perfect major accident prevention and mitigation strategy, make reactor core probability of damage CDF lower than 10
-6/ heap year, early stage a large amount of radiomaterials discharge probability LERF lower than 10
-7/ heap year.
Claims (25)
1. based on a non-active nuclear steam supply system of actively adding of 177 reactor cores, comprise nuclear reactor, reactor coolant loop, is characterized in that: described nuclear reactor comprises that 177 active lengths are the nuclear fuel assembly of 12 to 14 feet;
Described reactor coolant loop comprises reactor pressure vessel, main pump, steam generator, voltage stabilizer, release case and the coolant entrance of coupled reaction core pressure vessel and the main pipeline of outlet; Described main pipeline comprises cold section, hot arc and transition section, described hot arc is connected to primary side of steam generator entrance, described primary side of steam generator outlet is connected with transition section pipeline one end, and the other end of described transition section is communicated with cold section, thereby forms reactor-loop; Described voltage stabilizer lower end is connected on hot arc by fluctuation pipe; On described transition section and fluctuation pipe, be provided with LBB leakage detector.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described reactor coolant loop also comprises main pump, release case;
Described main pump is arranged on described cold section; Between the entrance and exit of described main pump, be provided with pressure tester; This voltage stabilizer top is provided with the first safe ozzle and the second safe ozzle, and described the first safe ozzle is connected with release main pipeline one end by the first pressure relief pipeline; The described release main line other end is communicated with release case portion; Described the second safe ozzle is connected on release main line by the second pressure relief pipeline; On described the second pressure relief pipeline, be provided with pressurizer safety valve; Described the first pressure relief pipeline and release main pipeline junction are provided with fast pressure relief valve, are provided with the valve of controlling flow on the first pressure relief pipeline between this fast decompression valve and described the first safe ozzle; Described reactor pressure vessel top is connected on described release main line by pressure vessel top gas exhaust piping; The vent valve of some series connection is set on the gas exhaust piping of described pressure vessel top.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described reactor pressure vessel volume is 50～80m
3.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described 177 fuel assemblies be arranged to 15 row, 15 row reactor cores; Each nuclear fuel assembly is square fuel assembly, and it comprises 264 nuclear fuel rods, 24 control rod guide pipes and the 1 measurement instrument pipe that is arranged in 17 row, l7 row.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described nuclear reactor thermal technology allowance is greater than 15%; The first circulation refulling cycle is 18～24 months.
According to claimed in claim 5 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: in the first circulation of reactor core, described 177 nuclear fuel assemblies comprise several burnable poison fuel assemblies, described burnable poison fuel assembly comprises some burnable poison fuel rods, described burnable poison fuel rod comprises several burnable poison fuel pellets, and described burnable poison fuel pellet is by Gd
2o
3with UO
2mix and sintering formation.
According to claimed in claim 6 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: Gd in described burnable poison fuel pellet
2o
3percentage by weight be 2%～10%.
According to claimed in claim 2 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described reactor pressure vessel, steam generator, main pump, voltage stabilizer, main pipeline designed life is 60 years.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described reactor pressure vessel top is also provided with the integrated heap top system of enclosing cylinder with missile roof shielding steel plate and lifting.
According to claimed in claim 1 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: also comprise that actively adding non-active heat discharges system, this actively adds non-active heat discharge system and comprises active system and passive system, and described active system comprises safety injection system, active Reactor cavity flooding system, safety shower system, steam generator auxiliary feedwater system; Described passive system comprises high-order safety injection system, non-active Reactor cavity flooding system, passive containment thermal conduction system, Passive residual heat removal system; Described safety injection system, active Reactor cavity flooding system, safety shower system, high-order safety injection system, non-active Reactor cavity flooding system are all communicated with reactor-loop by pipeline; Described steam generator auxiliary feedwater system, Passive residual heat removal system are all communicated with reactor secondary circuit by pipeline; Passive containment thermal conduction system is arranged on containment top.
11. according to claimed in claim 10 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described safety injection system comprises the first water source and the first active pipeline, and described first active pipeline one end is connected the other end with the first water source and is connected on reactor-loop.
12. according to described in claim 11 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described active Reactor cavity flooding system comprise second water source, one end be connected with second water source the other end be connected to reactor cavity bottom, one end connects the outside fire water supply other end and is connected to the fire extinguisher canvas hose on the second active pipeline.
13. according to described in claim 12 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described safety shower system comprises that the 3rd water source, the circular spray tube that is arranged on containment inner top, one end are connected with the 3rd water source the spray pipeline that the other end is connected with this circular spray tube.
14. according to described in claim 13 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described steam generator auxiliary feedwater system comprises that the 4th water source, one end are connected the 3rd active pipeline on the main feed water pipe road that the other end is connected to reactor secondary circuit with the 4th water source.
15. according to claimed in claim 10 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described high-order safety injection system comprises bottom the peace note case that is arranged on higher than reactor-loop position, one end and peace note case and is connected the first non-active pipeline that the other end is connected with cold section of reactor-loop.
16. according to claimed in claim 10 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described non-active Reactor cavity flooding system comprises that non-active Reactor cavity flooding case, one end of being arranged on higher than reactor core position are connected the second non-active pipeline that the other end is connected to reactor cavity bottom with non-active Reactor cavity flooding case.
17. according to claimed in claim 10 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described passive containment thermal conduction system comprises the first heat-exchanging water tank, is arranged on the second heat interchanger in containment, and described the second heat interchanger is communicated with the first heat-exchanging water tank by pipeline.
18. according to claimed in claim 10 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described Passive residual heat removal system comprises and be arranged on the second heat-exchanging water tank, be arranged on the 3rd heat interchanger in this second heat-exchanging water tank, the water inlet of described the 3rd heat interchanger and water delivering orifice are connected respectively on the main feed water pipe road and main steam line of reactor secondary circuit.
19. according to described in claim 14 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described the first water source, second water source, the 3rd water source are and are arranged on built-in material-changing water tank in containment vessel; Described the 4th water source is the auxiliary feed-water tank being arranged on outside containment vessel.
20. according to described in claim 14 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: also comprise instrument control system; Described instrument control system comprises reactor core measuring system, Protection of Diversity system DAS; Described reactor core measuring system comprises that several insert the detector assembly of reactor core from pressure vessel top, and described detector assembly comprises level sensor assembly and neutron-hygrosensor assembly; The signal output part of described level sensor assembly and neutron-hygrosensor assembly passes respectively containment and is connected to Core cooling monitor signal treatment facility and reactor core neutron flux signal handling equipment.
21. according to described in claim 20 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described level sensor assembly comprises inner two sensitive element M1 for hollow structure and M2; The inside of described M1 is provided with two K type thermopair HT1 and UHT1, and for the electric heater HE1 to HT1 heating; The inside of described M2 is provided with two K type thermopair HT2 and UHT2, and for the electric heater HE2 to HT2 heating; Described HT1 and HT2 are drive end, the reference edge that UHT2 is HT1, the reference edge that UHT1 is HT2; The inside of described sensitive element Ml and M2 is all filled with insulation course; The material of described insulation course is magnesium oxide.
22. according to described in claim 20 based on non-active nuclear steam supply system of actively adding of 177 reactor cores, it is characterized in that: described neutron-hygrosensor assembly comprises package shell, self-power neutron detector group, internal seal structure, connector, end part seal plug; Described self-power neutron detector group is fixed in described package shell, and it comprises N self-power neutron detector, a described N self-power neutron detector along described package shell axially from bottom to top evenly interval arrange; The vertical range of the lower surface, sensitive section of centre distance reactor fuel assemblies active region of i rhodium self-power neutron detector is from bottom to top
i=1 wherein, 2,3 ..., N; H is the axial height of reactor fuel assemblies active region; Described connector sealing is fixed on the upper end of described package shell; Described end part seal plug sealing is fixed on the lower end of described package shell; Described internal seal structure sealing is fixed on the inside of the described package shell that is positioned at described self-power neutron detector group top; The sheathed cable of described N rhodium self-power neutron detector is tied up and fixes, and axially extends upward along described package shell, through described internal seal structure, is finally fixed on the stitch of described connector; Also comprise and be fixed on the Pt100 four-wire system thermometer that is positioned at described internal seal structure upper area in described package shell, described Pt100 four-wire system thermometer comprises 4 lead-in wires; Described 4 lead-in wires are connected respectively on 4 stitch of described connector; Also comprise and be fixed on the thermopair that is positioned at described internal seal structure lower zone in described package shell; Described thermopair is connected on the stitch of described connector by thermopair extended line.
23. according to described in claim 20 based on non-active nuclear steam supply system of actively adding of 177 reactor cores; it is characterized in that: described Protection of Diversity system DAS adopts the equipment that is different from reactor protection system, to prevent the generation of the common mode failure identical with reactor protection system.
The nuclear power station of 24. 1 kinds, is characterized in that: adopt described in claim 1 to 22 any one based on non-active nuclear steam supply system of actively adding of 177 reactor cores;
Its power of the assembling unit 1000～1400MWe, average availability is more than or equal to 90%, and maximum ground acceleration is 0.3g, and containment is that double-deck steel design is clashed into Chinese People's Anti-Japanese Military and Political College's type business aircraft.
25. according to the nuclear power station described in claim 24, it is characterized in that: its nuclear reactor probability of damage CDF is lower than 10
-6/ heap year, early stage a large amount of radiomaterials discharge probability LERF lower than 10
-7/ heap year.
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| CN107507652A (en) * | 2017-07-31 | 2017-12-22 | 清华大学天津高端装备研究院 | The core structure and nuclear reactor of a kind of integral reactor |
| CN107507652B (en) * | 2017-07-31 | 2023-05-23 | 清华大学天津高端装备研究院 | Reactor core structure of integrated reactor and nuclear reactor |
| CN107393602B (en) * | 2017-07-31 | 2019-08-02 | 江苏华洋新思路能源装备股份有限公司 | A kind of passive fire resisting relief panel in nuclear power unit fuel treatment region |
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| CN108565034A (en) * | 2018-03-27 | 2018-09-21 | 中国核动力研究设计院 | U-tube flows backwards the system and method for thermal environment simulation under passive service condition |
| CN108899102A (en) * | 2018-07-06 | 2018-11-27 | 中国核动力研究设计院 | The method for arranging of the fixed self-power neutron detector general for presurized water reactor |
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| CN110111917A (en) * | 2019-04-17 | 2019-08-09 | 中广核工程有限公司 | Out-pile neutrons in nuclei flux monitoring method, apparatus and readable storage medium storing program for executing after accident |
| CN110415848A (en) * | 2019-08-05 | 2019-11-05 | 中国核动力研究设计院 | A kind of reply heat extraction reduces the protection system of superposition SWCCF accident |
| CN110415848B (en) * | 2019-08-05 | 2020-11-24 | 中国核动力研究设计院 | Protection system for reducing superimposed SWCCF accidents in response to heat extraction |
| CN111290443B (en) * | 2020-04-02 | 2023-04-07 | 广西防城港核电有限公司 | Continuous spraying flow control system of nuclear power station voltage stabilizer |
| CN111290443A (en) * | 2020-04-02 | 2020-06-16 | 广西防城港核电有限公司 | Continuous spraying flow control system of nuclear power station voltage stabilizer |
| CN112053791A (en) * | 2020-09-15 | 2020-12-08 | 哈尔滨工程大学 | Non-time-limit passive combined heat removal system with integrated heat release trap |
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| CN113782236A (en) * | 2021-09-09 | 2021-12-10 | 哈尔滨工程大学 | Test device for researching rapid start of nuclear reactor |
| CN113782236B (en) * | 2021-09-09 | 2024-01-16 | 哈尔滨工程大学 | Test device for researching quick start of nuclear reactor |
| CN114999693A (en) * | 2022-06-01 | 2022-09-02 | 中国核动力研究设计院 | Pressure relief protection system for preventing non-condensable gas from entering reactor core of compressed gas pressure stabilizing reactor |
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