CN101946253A - A methodology for modeling the fuel rod power distribution within a nuclear reactor core - Google Patents

A methodology for modeling the fuel rod power distribution within a nuclear reactor core Download PDF

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CN101946253A
CN101946253A CN2009801048522A CN200980104852A CN101946253A CN 101946253 A CN101946253 A CN 101946253A CN 2009801048522 A CN2009801048522 A CN 2009801048522A CN 200980104852 A CN200980104852 A CN 200980104852A CN 101946253 A CN101946253 A CN 101946253A
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fuel
rod
fuel rod
form factor
flux
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张宝成
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CBS Corp
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Westinghouse Electric Corp
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C5/00—Moderator or core structure; Selection of materials for use as moderator
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/001—Computer implemented control
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/001—Computer implemented control
    • G21D3/002—Core design; core simulations; core optimisation
    • 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
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • 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
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors

Abstract

A method for modeling a nuclear reactor core that follows the history of each fuel pin and employs fuel pin flux form factors to explicitly track each fuel pin's fluence exposure along its axial length and uses this information to obtain fundamental data for each fuel rod, i.e. fuel rod cross-sections, for each fuel pin segment. The data obtained for the fuel pins segments are employed to adjust the fuel pin flux form factors to match the real fuel pins' history so that the fuel rod power distribution can be precisely calculated based on the fuel rod cross-sections and the flux form factors.

Description

The method of the fuel rod distribute power in the simulation nuclear reactor
Technical field
The method of the distribute power in the relate generally to of the present invention simulation nuclear reactor relates more specifically to be designed for the method for the initial core and the reload core of nuclear reactor.
Background technology
Primary side with the water-cooled nuclear reactor electricity generation system of pressurization comprises the closed-loop path, and described closed-loop path isolates with the primary side that is used to produce useful energy and heat exchange relationship arranged with it.Primary side comprises: the reactor vessel of encapsulation core internal structure, a plurality of fuel assemblies that comprise fissile material of described core internal structure supporting; Primary return in the heat interchange steam generator; Be used to the to circulate inner volume of pressurizer, pump and pipeline of pressure (hydraulic) water, these pipelines are connected to reactor vessel independently with each steam generator and each pump.Comprise that each parts of primary side of the piping system, pump and the steam generator that are connected to container form the loop of primary side.Primary side also is connected to subsidiary loop, and described subsidiary loop comprises the cubing that is used for pressure (hydraulic) water and the loop of chemical monitoring.This subsidiary loop is arranged in the branch of primary return, make it possible to keep the water yield in the primary return by the water yield of additional survey in needs, and can monitor the chemical property of chilled water, especially for the content of the very important boric acid of the operation of reactor.
For illustrative purposes, Fig. 1 shows the nuclear reactor Entry-level System of simplification, and it comprises the reactor pressure vessel (10) of general cylindrical shape, and this reactor pressure vessel has the closure head (12) of encapsulation nuclear reactor (14).Be pumped in the container (10) by pump (16) by reactor core (14) and be disposed to heat exchanger (18) such as the liquid reactor coolant of water, heat energy is located to be absorbed at reactor core (14), this heat exchanger is commonly referred to as steam generator, and heat is passed to and utilizes the loop (not shown) such as the steam driven turbogenerator in described heat exchanger.Reactor coolant is got back to pump (16) then, finishes primary return.Typically, a plurality of aforesaid loops are connected to single reactor vessel (10) by reactor cooling duct (20).
Illustrate in greater detail exemplary reactor design among Fig. 2.Except comprising a plurality of reactor cores (14) parallel, the vertically collaborative fuel assembly (22) that extends, for this illustrative purposes, other container inner structure can be divided into inner structure (24) and last inner structure (26) down.In traditional design, the function of following inner structure is supporting, aims at and guiding reactor core parts and instrument, and in container direct flow.Last inner structure retrains or is provided for the secondary constraint of fuel assembly (22) (in the figure for easy, only illustrating two), and the instrument and the parts of supporting and guiding such as control rod (28).In the exemplary reactor shown in Fig. 2, cooling medium enters reactor vessel (10) by one or more inlet nozzles (30), flow downward by the annular space between container and the reactor core tube (32), turning over 180 ° in the pressure stabilizing chamber (34) down, upwards by lower support plate (37) and following core plate (36) (fuel assembly (22) is positioned on this time core plate (36)), and by near the fuel assembly.In some design, lower support plate (37) is replaced by the single structure (following core support plate) at the height place identical with (37) with following core plate (36).Flowing through reactor core and very big usually around the coolant flow in zone (38), is the magnitude of 400,000 gallons of per minutes under the speed of about 20 feet per seconds.Pressure drop that is produced and friction force are easy to make fuel assembly to rise, and this motion is by the last inner structure constraint that comprises circular upper core plate (40).The cooling medium that leaves reactor core (14) is along the following side flow of upper core plate and upwards by a plurality of perforation (42).Cooling medium upwards radially flows to one or more outlet nozzles (44) then.
Last inner structure (26) can and comprise upper support assembly (46) from the supporting of container or vessel head.Load is mainly transmitted between upper support assembly (46) and upper core plate (40) by a plurality of support columns (48).Perforation (42) top of support column in selected fuel assembly (22) and upper core plate (40) aimed at.
The spider assembly (52) and the driving shaft (50) that can straight-line control rod (28) typically comprise the neutron poison rod, described neutron poison rod is entered in the fuel assembly of having aimed at (22) by last inner structure (26) by control rod guide tube (54) guiding.Conduit is fixedly joined to upper support assembly (46) and is connected by the split pin in the top that is press fit into upper core plate (40) (56).Latch structure makes the conduit assembling and changes (if necessary) easily, and guarantees especially mainly to bear the reactor core load by support column (48) rather than conduit (54) under earthquake or the unexpected condition of other high load capacity.This helps to influence nocuously under the unexpected condition that control rod inserts ability in meeting and hinders deformation of guide tube.
Fig. 3 is a front elevation, and the form that shortens with vertical direction shows the fuel assembly of totally being represented by Reference numeral (22).Fuel assembly (22) is the type that is used in the pressurized water reactor, and has the structural framework that comprises bottom nozzle (58) in lower end.Bottom nozzle (58) is bearing in fuel assembly (22) in the following core support plate (60) in the (not shown) of the reactor core zone of nuclear reactor.Except bottom nozzle (58), the structural framework of fuel assembly (22) also comprises a plurality of conduits or guide bushings (54) and the top jet nozzle (62) that is positioned at the structural framework upper end, and these a plurality of conduits or guide bushings are longitudinally extended between bottom nozzle (58) and top jet nozzle (62) and its opposed end is attached to bottom nozzle (58) and top jet nozzle (62) rigidly.
Fuel assembly (22) also comprises and a plurality ofly axially separates and be installed to the horizontal grid spare (64) on the guide bushings (54) and laterally separated and the array of the organized microscler fuel rod (66) that supports by grid spare (64) along guide bushings (54).In addition, assembly (22) has the instrumentation tube (68) that is positioned at the center, and this instrumentation tube is extended between bottom nozzle (58) and top jet nozzle (62) and is installed to bottom nozzle (58) and top jet nozzle (62).Adopt such arrangements of components, fuel assembly (22) forms can be handled easily and can not damage the integral unit of component-assembled.
As mentioned above, the fuel rod (66) of the one-tenth array in the assembly (22) is by keeping with spaced relation along the isolated grid spare of fuel assembly length (64).Each fuel rod (66) all comprises fuel ball (70), and is sealed by upper end plug (72) and bottom plug (74) at the opposed end place.Pellet (70) piles up pressurized spring (76) between the top and is maintained in and piles up by being arranged on upper end plug (72) and pellet.The fuel pellet (70) that is made of fissile material is responsible for producing the reaction power of reactor.Specify in the assembly (22) fuel pellet (70) in the fuel rod (66) can be different aspect composition and enrichment with other fuel rod (66) in the same fuel assembly (22).Because the power of reactor output is subject to the maximum temperature that is stood along fuel rod (66), so the axially and radially distribute power of management reactor core is important.Operating conditions need be kept below the operating conditions that can cause departing from nuclear boiling along the covering of fuel rod (66).Under the condition of the type, the heat from fuel rod (66) to contiguous cooling medium is transmitted variation, and the temperature of fuel rod is risen, and this may cause covering to break down.Thereby dissimilar fuel rods is very important in the placement in the reactor core (14) in placement in the fuel assembly (22) and dissimilar fuel assemblies for guaranteeing security and maximization reactor core output efficiency.Liquid mitigator/cooling medium such as the water of water or boracic upwards is pumped to fuel assembly (22) by a plurality of flow openings in the following core support plate (60).The bottom nozzle (58) of fuel assembly (22) upwards passes through cooling medium through conduit (54) and along the fuel rod (66) of assembly, produce useful work so that absorb the heat that wherein produces.
In order to control fission process, a plurality of control rods (78) can be arranged in guide bushings (54) to-and-fro movement of fuel assembly (22) pre-position.Particularly, the cluster control gear (80) that is positioned at top jet nozzle (62) top supports this control rod (78).This control gear has tapped cylindrical hub parts (82), and this boss assembly has a plurality of anchor fluke or arms (52) that radially extend.Each arm (52) interconnects to control rod (78), make control rod mechanism (80) can operate to make control rod (78) motion vertically in guide bushings (54) under the dynamic action of control rod driving shaft (50), control the fission process in the fuel assembly (22) thus, all these is known mode.
As previously mentioned, the importantly design of the initial and reload core of management, so that the axially and radially distribute power of management reactor core, thereby guarantee security and maximize the reactor work efficiency.This means, must consider very carefully the placement of kind, these fuel rods of the interior fuel rod of assembly (22) (66) and assembly in reactor core placement so that the thermograde that occurs in the reactor core minimize.Current, the reactor core designs is for using the neutron diffusion coding, for example from the application's assignee Westinghouse Electric Company LLC, and the ANC that Pittsburgh, Pennsylvania secure permission.These neutron diffusion codings are divided into several energy ranges (energy bins) with neutron energy and are distributed by the core model estimating power.Since the geometric model of system intrinsic approximate with and the nuclear cross section database that adopted, so the precision of these estimations is not enough height.Advanced nodal method is used in current reactor core analytical calculation usually, and this method evenly turns to big node with the fuel pin in the fuel assembly (fuel pin), for example 17 takes advantage of 17 fuel stringer assembly to be converted into 2 * 2 nodal analysis method, as shown in Figure 4.For the nuclear reactor core that contains more than 100 fuel assemblies, next utilize this nodal analysis method to calculate three-dimensional neutron flux and distribute power.Based on reactor core wide node distribute power, by homogenising scheme and the combination of detailed form factor being generated the distribution of each fuel pin (being fuel rod) to assembly.Just can work well as long as simulated operation is historical clearly in the assembly that generates homogenising data and form factor calculates like this.Unfortunately, the true operation history of each fuel assembly was ignorant before this in advance, and this makes that being difficult to generate correct form factor comes accurately Simulation Core.
Under the reactor core operating conditions of reality, even for the fuel assembly (22) of same type, because surrounding environment (specifically being that control rod inserts and extract out historical), heterogeneity (being pointwise flux and distribute power) also will change during operation.In order to obtain the real history effect to the thin excellent power of fuel rod, prior art has attempted utilizing very complicated calculating to come thin excellent power form factor is carried out many kind corrections to generate the fuel assembly data.Yet the result is still very unsatisfactory, especially when the insertion of usually carrying out control rod or ash rod in normal power operating period with when extracting out.This designs in (for example current AP1000 that is provided by Westinghouse Electric Company LLC) at BWR reactor core design and new PWR reactor core becomes a big difficult problem.These problems do not produce owing to knowing in advance which assembly is control rod when, wherein will be inserted in.Be used for protecting history that the module data of reactor core design generates to be very different, and utilize traditional method in reactor core design coding, to be difficult to obtain this difference with the true fuel history that in reactor core, experiences during the normal running.
Therefore, need a kind of new method, it will predict the interior power and the Flux Distribution of reactor core of nuclear reactor better.
More specifically, need a kind of new method, it will be predicted under the situation of considering each fuel element in whole reactor core upper edge axially and radially power and Flux Distribution.
In addition, need a kind of new method, it will predict the distribute power on the reactor core of nuclear reactor better, and this distribute power reflects the history of reactor core more accurately.
In addition, need a kind of new method, it will predict the distribute power in the reactor core of nuclear reactor, and not need a large amount of computer processing times or storer.
Summary of the invention
Compare with traditional method, method of the present invention will be abolished thin excellent power form factor fully.On the contrary, method of the present invention is followed the exposure history of each fuel rod in reactor core, and draws fuel rod check figure certificate based on this real history, i.e. fuel pin cross section in the physical term (possibility of the neutron reaction of representative such as absorption, fission etc.).In actual applications, the real history of fuel rod is represented with fast fluence (fast fluence) by parametrization and by its burnup one by one.As most of reactor cores design coding, by carrying out from the fabricator to fuel rod power that current reactor core obtained simply and local neutron flux is calculated (drawing) these two parameters to the integration of time.In order to obtain the fuel pin cross section, under predetermined reactor operating conditions, normally under the total power level conditions of heat, generate in advance with reference to the cross section and show.For given natural fuel rod historical (burnup and fast fluence),, obtain the fuel pin cross section by inquiry cross section watch and with actual fluence and relatively carry out fast fluence correction with reference to fluence.During generation, also produce the thin excellent flux form factor of reference fuel with reference to thin excellent cross section table.The thin excellent flux form factor of the reference fuel that method of the present invention utilizes these to generate in advance in conjunction with above-mentioned fuel pin cross section, generates the actual thin excellent flux form factor that is used for given history.The real-time regulated of fuel pin flux form factor from the reference value to the actual value of utilizing the fuel pin cross section to carry out is based on the reactor physics basic theory of nuclear design encoding context.Therefore, above-mentioned fuel pin cross section and flux have been considered the history of fuel pin.Fuel pin unit cell Ka Bo fission (cell kappa-fission) is multiplied each other and will be provided the fuel pin distribute power with flux.
Description of drawings
From preferred embodiment explanation, can obtain further understanding of the present invention below in conjunction with accompanying drawing, wherein:
Fig. 1 is the rough schematic view that can use nuclear reactor system of the present invention;
Fig. 2 is the partial sectional view that can use nuclear reactor vessel of the present invention and internal part;
Fig. 3 is with the partial sectional view of the fuel assembly shown in the form that shortens vertically, has wherein for the sake of clarity removed the part parts;
Fig. 4 shows 2 * 2 nodal analysis methods that prior art adopts with figure;
Fig. 5 is the diagram of the part of fuel assembly, shows the individual difference in the fuel rod that the present invention considers; And
Fig. 6 is the process flow diagram of neoteric fuel rod power calculation.
Embodiment
In great majority nuclear reactor core design coding such as ANC, in order to obtain the thin excellent distribute power of each fuel rod, fuel pin power form factor is applied to the thin excellent power profile of homogenising on the whole node, so that obtain to be used for the thin one by one excellent homogenising distribute power of intranodal fuel assembly group.As many advanced persons' nuclear reactor core design coding, ANC uses the form factor that depends on energy bins.Just, the interior a plurality of fuel rods of a given cover form factor and given energy range are corresponding.Each energy bins (g) (x, the fuel rod of y) locating (thin rod) power is expressed as:
P g ( x , y ) = k Σ f , g hom ( x , y ) · φ g hom ( x , y ) · f g p ( x , y ) = P g hom ( x , y ) · f g p ( x , y ) - - - ( 1 )
Here,
Figure BPA00001197199400072
Be the thin excellent power of homogenising, it is by thin one by one excellent flux of homogenising and Ka Bo fission (k ∑ f, that is, come the energy release rate of self-fission) obtain.The thin excellent flux of homogenising
Figure BPA00001197199400073
Be to obtain by each individual node is separated two energy bins diffusion equations (flux of node side and node-angle) under the node boundary condition.As shown in Figure 4, each node is considered to single homogenising quality and supposes that the power form factor will can take into account all differences between fuel rod.Being used for each the Ka Bo fission of two energy bins of intranodal is the mean 0100 calorie pool fission of each fuel assembly in the corresponding energy bins, produces the mean value of 1.4061MeV/cm for energy bins 1, produces the mean value of 31.0616MeV/cm for energy bins 2.The Ka Bo fission of the thin rod one by one of homogenising
Figure BPA00001197199400074
The condition that is, side average from node and angle upper section utilizes polynomial expansion to generate, rather than utilizes true fuel rod state/history generation.
Adopt (the Ka Bo fission that x, the homogenising Ka Bo that y) locates fission and accurately do not represent corresponding fuel rod of this method.This method supposition will be by the power form factor
Figure BPA00001197199400075
Obtain heterogeneity (being the difference between the different fuel rod), the act as a fuel function of assembly average burn-up of this power form factor calculates (1attice code single-assembly calculation) by the trellis coding unimodule and generates in advance.
Different therewith, the method that is used for Simulation Core of the present invention is that the actual history (being fuel pin burnup and fast fluence) of each fuel pin of consideration is calculated the flux and the Ka Bo fission that are used for each fuel pin axial cross section section.Utilizing one by one, the heterogeneity Ka Bo fission and the flux of thin rod directly are calculated as fuel pin power:
P g ( x , y ) = k Σ f , g het ( x , y ) · φ g het ( x , y ) = k Σ f , g het ( x , y ) · φ g hom ( x , y ) · f g φ ( x , y ) - - - ( 2 )
Here,
Figure BPA00001197199400081
It is the fuel pin flux form factor.Similar with the power form factor, be used for each thin excellent reference flux form factor
Figure BPA00001197199400082
Calculate generation in advance by the trellis coding unimodule under predetermined condition, this predetermined condition is generally for example total power condition of heat.One group of fuel burn-up process from new life (0) to high combustion (for example 80MWD/kg) is selected as with reference to historical point.With reference to history point place, the trellis coding of each fuel section by being used for energy bins 1 and 2 calculates flux form factor at these.Each the exemplary card pool fission of sample fuel rod section that is used for energy bins 1 and 2 is shown in the following table:
Energy bins 1 Energy bins 2
0 0
1.38999 30.33812
1.40734 28.37248
1.41078 26.33429
1.42007 34.3432
1.41119 32.2942
1.40615 32.48921
Method of the present invention utilizes each fuel rod shown in Fig. 5 to carry out work, and wherein different shades is represented the difference between the fuel rod, i.e. difference in the fuel rod history, burnup etc. for example, and the difference in the excellent type, i.e. composition and enrichment.Thereby, represent each fuel rod by the cross section that equation 2 obtains.Based on the reactor physics theory, the fuel pin flux form factor depends primarily on the cross section of fuel pin one by one.Method of the present invention also adopts correction model based on reference and actual cross-section the fuel pin flux form factor to be adjusted to the thin rod spare of coupling natural fuel from the reference flux form factor, that is:
f g φ ( x , y ) = f g φ , ref ( x , y ) · F ( Σ ref , Σ act ) = f g φ , ref ( x , y ) · Π g ′ = 1 , g Σ a g ′ , ref · Σ g ′ → g act Σ a g ′ , act · Σ g ′ → g ref - - - ( 3 )
Here,
Figure BPA00001197199400084
And ∑ G ' → gRepresentative absorbs and disperses (from energy bins g ' to g) cross section respectively, and " ref " and " act " represents thin excellent cross section of reference fuel and the thin excellent cross section of natural fuel respectively.The process flow diagram of aforementioned process has been shown among Fig. 6.
The method of prior art is not considered the actual history of each single fuel rod.In other words, the actual inhomogeneity of its supposition fuel assembly depends primarily on assembly average loss history (burnup), and less depending on for obtaining the path (how reaching the there) that this history is taked.The work for most of PWR of prior art fuel pin power method is good, because above-mentioned supposition can be accepted for traditional pwr unit, this pwr unit can not move grey rod or control rod with full power operation and except device shutdown during the normal running usually on one's own initiative.
If move the ash rod or insert control rod in normal device operating period, for example in boiling water reactor and new pressurized water reactor, take place such as the AP1000 design, situation will be different fully so.Insert control rod and cause the heteropical remarkable change of assembly.Can calculate this moment impact that obtains to insert control rod by extra trellis coding.But this impact adds up along with fuel consumption.The heterogeneity of inserting with control rod during loss changes to be same as does not far from have the heterogeneity of control rod insertion to change.This has produced big problem to art methods, because when do not know, wherein and need to insert rod in the reactor core and do not know how long this rod will stop under which kind of condition.When the method according to this invention obtained the fuel pin cross section on single thin excellent basis, this problem no longer was big problem.
In order to improve the result of art methods, carried out large-scale research.The method that is adopted has generated the fuel pin power form factor of the many different control rod history that are used under different condition, and is made into form.Calculate even increase a large amount of trellis codings, art methods still can not provide the model of the satisfaction of the accurate core power distribution plan of forecasting institute under having ready conditions.This is because the thin excellent power form factor of using in calculating is not represented the true heterogeneity of fuel assembly.
Method of the present invention is directly handled each single fuel rod (thin rod).Method of the present invention is not used the assembly average loss, and is to use fuel rod burnup and frequency spectrum history (time integral of fluence, fast neutron level), and it is from the whole history calculating/accumulation that is fabricated onto current state, thereby obtains the fuel rod cross section.Two kinds of parameters of this of each fuel rod not only define the current state of fuel rod, have also reflected historical path.No matter fuel assembly history is how complicated, always method of the present invention can be based on these two kinds of parameter computing fuels rod cross sections (∑ for example aThe fuel rod absorption cross section, the k ∑ fThe fuel rod fission energy discharges the cross section), and the true heterogeneity of fuel rod cross section and fuel assembly is mated, this is because this method is followed the history of each fuel rod by following the tracks of above-mentioned two kinds of parameters.In addition, by the correction of equation 3, the fuel pin flux form factor is corresponding with the authentic component heterogeneity.Therefore, method of the present invention obtains the history of fuel assembly and each single fuel rod in time automatically.
Method of the present invention need not carried out history different and complexity and calculate during the fuel assembly data generate.Method of the present invention is followed time dependent fuel pin real history, and directly calculates thin excellent unit cell data (pin cell data, the data on the increment cross section) based on the fuel pin real history.Therefore, method of the present invention can be handled all types of control rods and careful incendivity absorber insertion and the scene of extracting out.
Different with the current full fuel pin calculating of studying in many National Laboratories and university, method of the present invention will directly not separated diffusion equation or the transport equation (NGM-next generation method) that is not used for each thin rod.On the contrary, method of the present invention adopts similar 1.5 groups method to regulate thin one by one excellent flux simply.Because do not need repetition and the connection of thin rod one by one, this method is more faster than NGM, has reproduced well simultaneously to transport the result.Compare with art methods, method of the present invention needs the considerably less computer processing unit time to increase.
In addition, (for example, can obtain thin excellent historical data (burnup and fluence) in ANC) at great majority design coding.Therefore, do not need to preserve any extra single thin excellent data.Single fuel pin information is lot of data.Preserve any extra single thin excellent data and will increase the disk demand significantly and influence coding efficiency, this is one of maximum problem for NGM.
As mentioned above, insert and extract out for the historical or careful incendivity absorber of the control rod of any kind, method of the present invention will be improved the prediction of thin excellent power.In addition, method of the present invention is calculated homogenising required single flux and cross section again.This means, if use this method, then can be dirt cheap and effective and efficient manner carry out homogenising again.Homogenising can be used for directly solving many unsolved reactor core problems again, for example assembly bending (assembly bow), the analysis of MOX/UO2 reactor core, is used for the power mistake estimation of reactor core perimeter component etc.
Though described specific embodiments of the invention in detail, it will be appreciated by those skilled in the art that according to whole instruction of the present disclosure and can carry out various changes and replacement to these details.Therefore, disclosed specific embodiment only is illustrative and nonrestrictive for scope of the present invention, and scope of the present invention is limited by the whole range of claims and any and whole equivalents thereof.

Claims (6)

  1. One kind simulate nuclear fuel assembly axially and the method for the excellent distribute power of radial fuel, may further comprise the steps:
    A) neutron energy with described fuel assembly is divided into a plurality of energy bins;
    B) on a plurality of axial increment, consider each fuel rod in the fuel assembly individually;
    C), calculate the Ka Bo fission and the neutron flux form factor that are used for each fuel rod based on a plurality of reference values and actual fuel rod and fuel assembly history;
    D) generation is used for the homogenising neutron flux value of the fuel rod of described fuel assembly adjacent set;
    E) calculate a plurality of non-homogeneous neutron flux of each fuel rod in the fuel rod of described adjacent set respectively by homogenising flux and flux form factor;
    F) be identified for the power of each fuel rod by the constant multiple of the sum of products of the Ka Bo that calculated in all described a plurality of energy bins fission and non-homogeneous neutron flux.
  2. 2. method according to claim 1 is characterized in that, described flux form factor is the composition of fuel rod and the function of enrichment.
  3. 3. method according to claim 1, it is characterized in that, being identified for each fuel rod at first is the reference flux form factor with each historical flux form factor of analog fuel rod, and described reference flux form factor is not considered the placement history of described fuel assembly in reactor core.
  4. 4. method according to claim 3 comprises step: regulate described reference flux form factor, insert supposition history in the described fuel assembly to the influence of fuel rod burnup to consider control rod.
  5. 5. method according to claim 4 is characterized in that, described control rod inserts in the described fuel assembly during being assumed to be at the certain percentage of fuel assembly operation cycle.
  6. 6. method according to claim 3, it is characterized in that, history before the fuel assembly has been considered in the adjusting of described flux form factor, and the history before described comprises placement, partial power and the burnup that is experienced in the loaded cycle before each fuel assembly is in reactor core.
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CN103503076B (en) * 2011-03-15 2017-02-15 阿海珐核能公司 Method for operating a pressurized water reactor during load monitoring
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