CN101620892A - Structural design of loop system of high-power pressurized water reactor nuclear power station - Google Patents
Structural design of loop system of high-power pressurized water reactor nuclear power station Download PDFInfo
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- CN101620892A CN101620892A CN200910089130A CN200910089130A CN101620892A CN 101620892 A CN101620892 A CN 101620892A CN 200910089130 A CN200910089130 A CN 200910089130A CN 200910089130 A CN200910089130 A CN 200910089130A CN 101620892 A CN101620892 A CN 101620892A
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Abstract
The invention discloses a structural design of a loop system of a high-power pressurized water reactor nuclear power station, belonging to the technical field of nuclear power energy sources. The structural design comprises a main container and two bidirectional flow heat exchange tube pressurized water reactor steam generators, wherein the main container is provided with four inlets and four outlets; the two steam generators are adopted, each steam generator is provided with two inlets and two outlets, the cooling water pumps of four reactors are respectively connected with the four outlets of the two steam generators and then are connected with the inlets of the main container by cold section main pipelines, and the four outlets of the main container are respectively connected with the four inlets of the two steam generators by four hot section main pipelines, thereby ensuring the bidirectional flow circulation and the heat transmission of cooling water in the loop system. The invention is favorable to realizing the uniformity of a flow field of an upper cavity of a reactor core of a single high-power nuclear generator group in operation and lowering the stress level at the connection tubes of the outlets of the main container, and well solves the problem of higher safety and maintenance cost existing in the power maximization of a single reactor of the pressurized water reactor nuclear generator group.
Description
Technical field
The invention belongs to the nuclear power energy technology field, particularly a kind of high-power pressurized water reactor nuclear power station one circuit system structural design.
Background technology
The major equipment of pressurized-water reactor nuclear power plant one circuit system comprises: primary tank, steam generator, cooling-water pump, and the trunk line that connects these equipment.Present labour or the arrangement of a circuit system of building pressurized-water reactor nuclear power plant be divided into two generations four class.Wherein two classes, the one loop arrangement of second generation pressurized-water reactor nuclear power plant has: primary tank has the import of two symmetrical distributions and the outlet of two symmetrical distributions, each outlet links to each other with the inlet of a steam generator by a hot arc trunk line respectively, each primary tank import links to each other with cooling-water pump, steam generator outlet successively by a cold section trunk line respectively, constitutes two loop arrangements of two steam generators of primary tank two imports, two outlets with this; Second class is that primary tank has three equally distributed imports and three equally distributed outlets, per two adjacent imports link to each other with a steam generator by one cold section, cooling-water pump, a hot arc trunk line with outlet, constitute three three loop arrangements that export three steam generators of three imports of primary tank with this.The representative loop method for arranging of third generation pressurized-water reactor nuclear power plant mainly contains two classes: primary tank has four outlets of four imports (shown in Fig. 1 b), adjacent import links to each other with a steam generator by one cold section, cooling-water pump, a hot arc trunk line with outlet, constitutes four Fourth Ring road arrangements that export four steam generators of four imports of primary tank with this; An another kind of loop arrangement rough schematic view that is China shown in Figure 1 in the first core power station of building, wherein a. is the loop arrangement rough schematic view of nuclear power station AP1000.B is a loop arrangement rough schematic view of N4 nuclear power station.Number in the figure 100 is a primary tank, and label 101 is a steam generator among Fig. 1 a, Fig. 1 b, and Fig. 1 a is that 2, Fig. 1 b are 4.Label 109 is the heat pipe section of a circuit system among Fig. 1 a, Fig. 1 b, and Fig. 1 a is that 2, Fig. 1 b are 4.Label 107,108,110 and 111 is the cold leg of a circuit system among Fig. 1 a, Fig. 1 b, and promptly primary tank 100 has four to be 107,108,110 circuit system imports that are connected with four cold legs of 111 with label; Steam generator adopts △-125 type steam generator; Label is that four cooling-water pumps of 103,104,105 and 106 are connected to a circuit system import by 107,108,110 and 111 4 cold legs, realizes the chilled water circulation of a circuit system.
If the presurized water reactor single-machine capacity of the N4 nuclear power station of the above-mentioned AP1000 nuclear power station that has only two discharge connections or four discharge connections is to the better large pressurized water reactor development of economy, the unevenness in primary tank upper chamber flow field will further be aggravated, and might bring the stream at upper chamber's inner structural members such as control rod guide cylinder and discharge connection place to cause vibration problem; Therefore, China build with active service nuclear power station unit than large-scale advanced pressurized water reactor unit, power and economy still remain to be improved, and their a loop arrangement can not satisfy the security of more high-power nuclear power station one loop down layout and the requirement of economy; And external existing large nuclear power station research approach belongs to third generation moderate improvement type advanced pressurized water reactor nuclear power plant basically, and its economy and security still have the leeway of further raising.
Summary of the invention
The purpose of this invention is to provide a kind of high-power pressurized water reactor nuclear power station one circuit system structural design, it is characterized in that described high-power pressurized water reactor nuclear power station one circuit system comprises primary tank and two the two-way flow heat exchanger tube PWR steam generators with four imports, four outlets; Adopt every two steam generators that two inlets and two outlets are arranged, the cooling-water pump of 4 reactors links to each other with four outlets of two steam generators respectively, the inlet of logical again super cooled sect trunk line of cooling-water pump and primary tank joins, and four outlets of primary tank are joined by four inlets of four hot arc trunk lines and described two steam generators respectively, thereby have guaranteed the two-way flow circulation and the heat delivered of chilled water in the circuit system.
Four imports and four outlets of described primary tank upwards are equally distributed in the week of primary tank, and on each comfortable same surface level; The axial height at four outlet places will be lower than the axial height at four import places.
Described steam generator is primary side four chamber two-way flow heat exchanger tube PWR steam generators, " ten " font dividing plate is divided into four hydroeciums that cold and hot hydroecium replaces mutually with hydroecium in the low head hydroecium, connect the tube bank of inverted U pipe at low head hydroecium tube sheet upper surface, in installation cooling water inlet and the outlet corresponding of low head hydroecium bottom with hot hydroecium and cold water chamber.
Four imports of described primary tank and the cross-sectional area of four outlets equate.
The invention has the beneficial effects as follows in the reliability of having taken all factors into consideration nuclear power station, economy, use for reference the basis of the loop arrangement of 4 loops, 4 steam generators and AP1000, AP1000 loop under the high-power compressed water reactor nuclear power unit method of arranging more helps realizing the homogeneity in the reactor core upper chamber flow field of primary tank under 4 imports and 4 exit conditions, and under the identical primary tank outlet total area condition, reduce the stress level at primary tank discharge connection place; Also having reduced return building materials and factory building takes up an area of.
Description of drawings
Fig. 1 is the loop arrangement rough schematic view of the first core power station AP1000 of prior art, and wherein a. is the loop arrangement rough schematic view of nuclear power station AP1000; B. be a loop arrangement rough schematic view of N4 nuclear power station.
Fig. 2 is the circuit system structural representation of primary tank four imports, four outlets of the present invention and two steam generators.Wherein (a) is circuit system structural arrangement mode synoptic diagram; (b) be circuit system structural arrangement mode rough schematic view.
Fig. 3 is the right steam generator synoptic diagram of Fig. 2.
Fig. 4 is the left steam generator synoptic diagram of Fig. 2.
Fig. 5 is the upper chamber's velocity field under Fig. 1 scheme.
Fig. 6 is the upper chamber's velocity field under Fig. 2 scheme.
Embodiment
The invention provides a kind of high-power pressurized water reactor nuclear power station one circuit system.Below in conjunction with embodiment and accompanying drawing thereof this present invention is further described.
Shown in Figure 2 is the circuit system structural representation of primary tank four imports, four outlets of the present invention and two steam generators.Wherein (a) is circuit system structural arrangement mode synoptic diagram; (b) be circuit system structural arrangement mode rough schematic view.Among the figure, on upper chamber 2 positions of primary tank 1, along being uniform-distribution with first cooling water inlet 6 on the circumferential direction, second cooling water inlet 9, the 3rd cooling water inlet 13 and the 4th cooling water inlet 18, and first coolant outlet 5, second coolant outlet 10, the 3rd coolant outlet 14 and the 4th coolant outlet 17, wherein first cooling water inlet 6, second cooling water inlet 9, the surface level at the 3rd cooling water inlet 13 and 18 places, the 4th cooling water inlet will be higher than first coolant outlet 5 of primary tank in the axial direction, second coolant outlet 10, the surface level at the 3rd coolant outlet 14 and the 4th coolant outlet 17 places; First coolant outlet 5, second coolant outlet 10, the 3rd coolant outlet 14 and the 4th coolant outlet 17 respectively with the first hot arc trunk line 7 of primary tank, the second hot arc trunk line 11, the 3rd hot arc trunk line 16, the 4th hot arc trunk line 19 links to each other, the first hot arc trunk line 7, the other end of the second hot arc trunk line 11 enters the mouth 22 with first of right steam generator 4 respectively, second inlet 24 is connected, the 3rd hot arc trunk line 16, the other end of the 4th hot arc trunk line 19 enters the mouth 26 with the 3rd of left steam generator 3 respectively, the 4th inlet 28 is connected, first cold section trunk line 8 of primary tank, secondary cooling zone trunk line 12, the 3rd cold section trunk line 15, one end of the 4th cold section trunk line 20 respectively with first cooling-water pump 29, second cooling-water pump 30, the 3rd cooling-water pump 31, the discharge connection of the 4th cooling-water pump 32 links to each other, first cooling-water pump 29, the inlet connection of second cooling-water pump 30 exports 21 with first of right steam generator 4 respectively, second outlet 23 links to each other the 3rd cooling-water pump 31, the inlet connection of the 4th cooling-water pump 32 exports 25 with the 3rd of left steam generator 3 respectively, the 4th outlet 27 links to each other.Four imports of above-mentioned cold section trunk line, hot arc trunk line and primary tank and the cross-sectional area of four outlets equate.
Fig. 3 is the right steam generator synoptic diagram of Fig. 2, steam generator shown in the figure is primary side four chamber two-way flow heat exchanger tube PWR steam generators, in low head hydroecium 303, " ten " font dividing plate 316 is divided into first hot hydroecium 304 and four hydroeciums such as first cold water chamber 310, second hot hydroecium 305 and second cold water chamber 311 that cold and hot hydroecium replaces mutually with hydroecium; The inverted U pipe that connects arrangement in order at low head hydroecium tube sheet 303 upper surfaces restrains 307, and installation chilled water first inlet 21, second inlet 23 and first outlet 22, second corresponding with hot hydroecium and cold water chamber in low head hydroecium 303 bottoms exports 24.
Fig. 4 is the left steam generator synoptic diagram of Fig. 2.Left side steam generator 3 structures and right steam generator 4 structures are identical, and just the parts label changes: the inlet connection of the 3rd cooling-water pump 31, the 4th cooling-water pump 32 links to each other with the 3rd outlet the 25, the 4th outlet 27 of left steam generator 3 respectively.
When a circuit system of arranging moves according to the method described above, the chilled water of High Temperature High Pressure is respectively from first coolant outlet 5 of the primary tank 1 of reactor, second coolant outlet 10, the 3rd coolant outlet 14 and the 4th coolant outlet 17 enter the first hot arc trunk line 7 of primary tank, the second hot arc trunk line 11, the 3rd hot arc trunk line 16, the 4th hot arc trunk line 19, the first hot arc trunk line 7 wherein, chilled water in the second hot arc trunk line 11 enters chilled water first inlet 21 of right steam generator 4, second inlet, 23, the three hot arc trunk lines 16, chilled water in the 4th hot arc trunk line 19 enters the 3rd inlet 26 of left steam generator 3, the 4th inlet 28; After the chilled water of High Temperature High Pressure cooled off in left steam generator 3, right steam generator 4, first outlet, 22, second outlet 24 by right steam generator 4 entered first cooling-water pump 29, second cooling-water pump 30 respectively; The 3rd outlet 25 by left steam generator 3, the 4th outlet 27 enters the 3rd cooling-water pump 31, the 4th cooling-water pump 32, first cooling-water pump 29 afterwards, second cooling-water pump 30 pumps into first cooling water inlet 6 with primary tank 1 with cooled chilled water, first cold section trunk line 8 that second cooling water inlet 9 links to each other, secondary cooling zone trunk line 12, go forward side by side and continue circulation, the 3rd cooling-water pump 31 into primary tank 1, the 4th cooling-water pump 32 also pumps into cooled chilled water the 3rd cold section trunk line 15 that links to each other with the 4th cooling water inlet 18 with the 3rd cooling water inlet 13 of primary tank 1, the 4th cold section trunk line 20 enters next circulation after entering the chilled water that primary tank 1 continues to be heated to be High Temperature High Pressure.
Fig. 5 is under the outlet total area initial physical parameter equal, chilled water the design conditions identical with total flow with Fig. 6, velocity distribution situation in the upper chamber flow field under two kinds of contrast schemes, contrast shows, upper chamber of the present invention Flow Field Distribution is more even, more helps the safety of reactor core upper chamber inner structural members.
Claims (5)
1. high-power pressurized water reactor nuclear power station one a circuit system structural design is characterized in that, described high-power pressurized water reactor nuclear power station one circuit system comprises primary tank and two the two-way flow heat exchanger tube PWR steam generators with four imports, four outlets; Adopt every two steam generators that two inlets and two outlets are arranged, the cooling-water pump of 4 reactors links to each other with four outlets of two steam generators respectively, the inlet of logical again super cooled sect trunk line of cooling-water pump and primary tank joins, and four outlets of primary tank are joined by four inlets of four hot arc trunk lines and described two steam generators respectively, thereby have guaranteed the two-way flow circulation and the heat delivered of chilled water in the circuit system.
2. according to the described a kind of high-power pressurized water reactor nuclear power station one circuit system structural design of claim 1, it is characterized in that four imports and four outlets of described primary tank upwards are equally distributed in the week of primary tank, and on each comfortable same surface level; The axial height at four outlet places will be lower than the axial height at four import places.
3. according to the described a kind of high-power pressurized water reactor nuclear power station one circuit system structural design of claim 1, it is characterized in that, described steam generator is primary side four chamber two-way flow heat exchanger tube PWR steam generators, " ten " font dividing plate is divided into cold and hot hydroecium four hydroeciums alternately mutually with hydroecium in the low head hydroecium, connect the tube bank of inverted U pipe at low head hydroecium tube sheet upper surface, in installation cooling water inlet and the outlet corresponding of low head hydroecium bottom with hot hydroecium and cold water chamber.
4. according to the described a kind of high-power pressurized water reactor nuclear power station one circuit system structural design of claim 1, it is characterized in that four imports of described primary tank and the cross-sectional area of four outlets equate.
5. according to the described a kind of high-power pressurized water reactor nuclear power station one circuit system structural design of claim 1, it is characterized in that the cross-sectional area of described cold section trunk line and hot arc trunk line equates.
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| CN2009100891306A CN101620892B (en) | 2009-07-30 | 2009-07-30 | Structural design of loop system of high-power pressurized water reactor nuclear power station |
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| CN2009100891306A CN101620892B (en) | 2009-07-30 | 2009-07-30 | Structural design of loop system of high-power pressurized water reactor nuclear power station |
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| CN101839467A (en) * | 2010-05-10 | 2010-09-22 | 中国核电工程有限公司 | Method for mounting main pipeline and main loop of pressurized water reactor nuclear power station steam generator |
| CN102324205A (en) * | 2011-08-30 | 2012-01-18 | 华北电力大学 | Experimentally runnable dynamic simulation model for pressurized water reactor and steam generator thereof |
| CN104197307A (en) * | 2014-08-25 | 2014-12-10 | 华北电力大学 | Steam generator supporting partition plate structure with efficient heat transfer performance |
| CN104992738A (en) * | 2014-07-29 | 2015-10-21 | 台山核电合营有限公司 | Nuclear island main equipment installation process based on three-dimensional high-precision measurement |
| CN105913889A (en) * | 2016-07-05 | 2016-08-31 | 上海核工程研究设计院 | Three-loop nuclear energy system |
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| US4769209A (en) * | 1986-01-10 | 1988-09-06 | Westinghouse Electric Corp. | Compact small pressurized water nuclear power plant |
| FR2614462B1 (en) * | 1987-04-27 | 1989-07-28 | Framatome Sa | METHOD FOR COMPLETELY REPLACING A STEAM GENERATOR OF A PRESSURE WATER NUCLEAR REACTOR |
| US5263067A (en) * | 1992-06-24 | 1993-11-16 | Westinghouse Electric Corp. | Locating hot and cold-legs in a nuclear powered steam generation system |
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| CN101839467A (en) * | 2010-05-10 | 2010-09-22 | 中国核电工程有限公司 | Method for mounting main pipeline and main loop of pressurized water reactor nuclear power station steam generator |
| CN102324205A (en) * | 2011-08-30 | 2012-01-18 | 华北电力大学 | Experimentally runnable dynamic simulation model for pressurized water reactor and steam generator thereof |
| CN104992738A (en) * | 2014-07-29 | 2015-10-21 | 台山核电合营有限公司 | Nuclear island main equipment installation process based on three-dimensional high-precision measurement |
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