EP2240939A2 - 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 coreInfo
- Publication number
- EP2240939A2 EP2240939A2 EP09739300A EP09739300A EP2240939A2 EP 2240939 A2 EP2240939 A2 EP 2240939A2 EP 09739300 A EP09739300 A EP 09739300A EP 09739300 A EP09739300 A EP 09739300A EP 2240939 A2 EP2240939 A2 EP 2240939A2
- Authority
- EP
- European Patent Office
- Prior art keywords
- fuel
- pin
- history
- fuel rod
- rod
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Withdrawn
Links
Classifications
-
- 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
Definitions
- the present invention relates generally to a method of modeling the power distribution within the core of a nuclear reactor and more particularly to a method for designing initial and reload cores for a nuclear reactor.
- the primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated and in heat-exchange relationship with a secondary side for the production of useful energy.
- the primary side comprises the reactor vessel enclosing a core internals structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently.
- Each of the parts of the primary side comprising a steam generator, a pump and a system of pipes which are connected to the vessel form a loop of the primary side.
- Fig. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel (10) having a closure head (12) enclosing a nuclear core (14).
- a liquid reactor coolant such as water is pumped into the vessel (10) by pump 16 through the core (14) where heat energy is absorbed and is discharged to a heat exchanger (18), typically referred to a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator.
- the reactor coolant is then returned to the pump (16), completing the primary loop.
- a plurality of the above described loops arc connected to a single reactor vessel (10) by reactor cooling piping (20).
- the other vessel internal structures can be divided into the lower internals (24) and the upper internals (26).
- the lower internals function is to support, align and guide core components and instrumentation, as well as direct flow within the vessel.
- the upper internals restrain or provide a secondary restraint for the fuel assemblies (22) (only two of which are shown for simplicity in this figure), and support and guide instrumentation and components, such as control rods (28).
- the fuel assemblies (22) only two of which are shown for simplicity in this figure
- support and guide instrumentation and components such as control rods (28).
- coolant enters the reactor vessel (10) through one or more inlet nozzles (30), flows down through an annulus between the vessel and the core barrel (32), is turned 180° in a lower plenum (34), passes upwardly through a lower support plate (37) and a lower core plate (36) upon which the fuel assemblies (22) are seated and through and about the assemblies.
- the lower support plate (37) and the lower core plate (36) are replaced by a single structure, the lower core support plate, at the same elevation as (37).
- the coolant flow through the core and surrounding area (38) is typically large, on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second.
- Coolant exiting the core (14) flows along the underside of the upper core plate and upwardly through a plurality of perforations (42). The coolant then flows upwardly and radially to one or more outlet nozzles (44).
- the upper internals (26) can be supported from the vessel or the vessel head and include an upper support assembly (46). Loads are transmitted between the upper support assembly (46) and the upper core plate (40), primarily by a plurality of support columns (48). A support column is aligned above a selected fuel assembly (22) and perforations (42) in the upper core plate (40).
- Rectilinearly moveable control rods (28) typically include a drive shaft (50) and a spider assembly (52) of neutron poison rods that are guided through the upper internals (26) and into aligned fuel assemblies (22) by control rod guide tubes (54).
- the guide tubes are fixedly joined to the upper support assembly (46) and connected by a split pin (56) forced fit into the top of the upper core plate (40).
- the pin configuration provides for ease of guide tube assembly and replacement if ever necessary and assures that core loads, particularly under seismic or other high loading accident conditions are taken primarily by the support columns (48) and not the guide tubes (54). This assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability.
- FIG. 3 is an elevational view, represented in vertically shortened form, of a fuel assembly being generally designated by reference character (22).
- the fuel assembly (22) is the type used in a pressurized water reactor and has a structural skeleton which, at its lower end includes a bottom nozzle (58).
- the bottom nozzle (58) supports the fuel assembly (22) on a lower core support plate (60) in the core region of the nuclear reactor (not shown).
- the structural skeleton of the fuel assembly (22) also includes a top nozzle (62) at its upper end and a number of guide tubes or thimbles (54), which extend longitudinally between the bottom and top nozzles (58) and (62) and at opposite ends are rigidly attached thereto.
- the fuel assembly (22) further includes a plurality of transverse grids (64) axially-spaced along, and mounted to the guide thimbles (54) and an organized array of elongated fuel rods (66) traversely-spaced and supported by the grids (64). Also, the assembly (22) has an instrumentation tube (68) located in the center thereof and extending between, and mounted to, the bottom and top nozzles (58) and (62). With such an arrangement of parts, fuel assembly (22) forms an integral unit capable of being conveniently handled without damaging the assembly of parts. [0009] As mentioned above, the fuel rods (66) in the array therof in the assembly (22) are held in spaced relationship with one another by the grids (64) spaced along the fuel assembly length.
- Each fuel rod (66) includes nuclear fuel pellets (70) and is closed at its opposite ends by upper and lower end plugs (72) and (74).
- the pellets (70) are maintained in a stack by a plenum spring (76) disposed between the upper end plug (72) and the top of the pellet stack.
- the fuel pellets (70), composed of fissile material, are responsible for creating the reactive power of the reactor.
- the fuel pellets (70) within a given fuel rod (66) within an assembly (22) may vary in composition and enrichment from other fuel rods (66) within the same fuel assembly (22). It is important to manage the axial and radial power profile of the core because the power output of the reactor is limited by the hottest temperature experienced along a fuel rod (66).
- a liquid moderator/coolant such as water or water containing boron, is pumped upwardly through a plurality of flow openings in the lower core support plates (60) to the fuel assembly (22).
- the bottom nozzle (58) of the fuel assembly (22) passes the coolant upwardly through the guide tubes (54) and along the fuel rods (66) of the assembly in order to extract heat generated therein for the production of useful work.
- a number of control rods (78) are reciprocally moveable in the guide thimbles (54) located at predetermined positions in the fuel assembly (22).
- a rod cluster control mechanism (80) positioned above the top nozzle (62) supports the control rods (78).
- the control mechanism has an internally threaded cylindrical hub member (82) with a plurality of radially-extending flukes or arms (52).
- Each arm (52) is interconnected to the control rod (78) such that the control rod mechanism (80) is operable to move the control rods (78) vertically in the guide thimbles (54) to thereby control the fission process in the fuel assembly (22), under the motive power of control rod drive shafts (50) which are coupled to the control rod hubs (80), all in a well known manner.
- the method of this invention will completely do away with the pin power form factors. Instead, the method of this invention follows the exposure history of each fuel rod in the core, and, based on that real history, derives the fuel rod nuclear data, i.e. fuel pin cross-sections in a physics terminology (to represent the probability of neutron reaction such as absorption, fission, etc.).
- the fuel rod by rod true history is parameterized and represented by its burnup along with the fast fluence. As with most core design codes, these two parameters are calculated (followed) by simply doing a time integration of fuel rod power and local neutron flux obtained from the manufacturer to the current core.
- a reference cross-section table is pre-generated at a pre-defined reactor operating condition, typically the hot full power level condition.
- a pre-defined reactor operating condition typically the hot full power level condition.
- the fuel pin by fuel pin cross sections are derived through looking at the cross-section table and performing a fast fluence correction by comparing actual fluence with the reference one.
- a reference fuel pin flux form factor table is created as well. The method of this invention uses these pre-generated reference fuel pin flux form factors, in conjunction with the above fuel pin by fuel pin cross sections, to generate the actual pin flux form factors for the given history.
- FIG. 1 is a simplified schematic of a nuclear reactor system to which this invention may be applied;
- Fig. 2 is an elevational view, partially in section, of a nuclear reactor vessel and internal components to which this invention may be applied;
- Fig. 3 is an elevational view, partially in section, of a fuel assembly illustrated in vertically shortened form, with parts broken away for clarity;
- Fig. 4 graphically illustrate a 2 x 2 nodal model employed by the prior art;
- Fig. 5 is a graphical representation of a portion of the fuel assembly illustrating the individual differences in fuel rods taken into account by this invention.
- Fig. 6 illustrates the flowchart of the new invention on fuel rod power calculation.
- P ⁇ om ⁇ x,y is the homogeneous pin power, which is obtained from the homogeneous pin-by-pin fluxes and kappa-fissions ( ⁇ f , i.e. energy release rate from fission).
- the homogeneous pin fluxes ⁇ g Om (x,y) are derived by solving two energy group diffusion equations for each individual node along with the node boundary conditions (node sides' and corners' fluxes). Each node, as illustrated in Fig. 4 is considered as a single homogeneous mass and assumes the power form factors will take care of all the differences among the fuel rods.
- the kappa- fissions for each of the two energy groups within the node are the average kappa-fissions of each of the fuel assemblies within the corresponding energy group yielding an average value of 1.4061 MeV/cm for Energy Group 1 and 31.0616 MeV/cm for Energy Group 2.
- the homogeneous pin-by-pin kappa-fissions x ⁇ f imm (x,y) are generated using a polynomial expansion from the conditions of nodal average, sides, and corner cross- sections, rather than the real fuel rod status/history.
- the homogeneous kappa-fissions at (x,y) employing this method do not accurately represent the kappa-fissions of the corresponding fuel rod.
- the method assumes that the heterogeneity, i.e., the difference among different fuel rods, will be captured by the power form factors ff ⁇ x,y) , which, as a function of fuel assembly average burnup, are generated in advance through lattice code single-assembly calculations.
- the method of this invention for modeling a core calculates the fluxes and the kappa-fissions for each fuel pin axial cross-section segments taken into account the actual history of each fuel pin, i.e. fuel pin burnup and fast fluence.
- f*(x,y) is the fuel pin flux form factor. Similar to the power form factor, the reference flux form factor, f*[x,y) for each pin is generated in advance through lattice code single-assembly calculations under pre-defined conditions, e.g. typically a hot full power condition. A set of fuel burnup steps from fresh (0) to high burned (for instance 80MWD/kg) are chosen as the reference history points. At these reference history points, the flux form factors are calculated through a lattice code for each of the fuel segments for Energy Groups 1 and 2. Exemplary kappa-fissions for sample fuel rod cross-section segments in each of the Energy Groups 1 and 2 are shown in the chart below:
- the method of this invention works with each fuel rod as shown in Fig. 5, with the different shades representing the differences between fuel rods, i.e., difference in fuel rod history, e.g. burnup, etc. and difference in type of rod, i.e., composition and enrichment.
- the cross-section obtained from Equation 2 represents each fuel rod.
- the fuel pin flux form factor mainly depends on fuel pin by pin cross-sections.
- the method of this invention also adopts a correction model to adjust the fuel pin flux form factor from the reference flux form factor to match the actual fuel pin condition based on the reference and actual cross sections, i.e.:
- ⁇ * and ⁇ g - ⁇ g stand for absorption and scattering (from energy group g' to g) cross-sections respectively, and "ref ' and "act” for reference and actual fuel pin cross- sections.
- Fig. 6 A flowchart of the foregoing process is illustrated in Fig. 6.
- the method of the prior art does not take into account the actual history of each individual fuel rod.
- the actual heterogeneity of a fuel assembly mainly depends on the assembly average depletion history (burnup) and is less dependent on the path that was taken to obtain that history (how it got there).
- the prior art fuel pin power method works fine for most PWR's since the above assumption is acceptable for conventional PWR plants that typically run at full power during normal operation and do not actively move gray rods or control rods except for plant shutdown.
- control rod insertion leads to a significant change in the assembly heterogeneity. This instantaneous impact of control rod insertion can be captured through additional lattice code calculations. But this impact is accumulated with the fuel depletion. The heterogeneity change with control rod insertion during depletion is far different from that without control rod insertion. This creates a big issue for the prior art method since we don't know when, where, and under what conditions the rods need to be inserted into the core and how long they will stay. This issue is not as big a problem when the fuel pin cross-sections are taken on an individual pin basis in accordance with the method of this invention.
- the method of this invention uses fuel rod burnup and spectrum history (fluence, the time integration of fast neutron level) which arc calculated/accumulated over the history from manufacture to the current state to get the fuel rod cross-sections.
- These two parameters of each fuel rod define not only the fuel rod's current state but also reflect the path of the history.
- the method of this invention is always able to calculate the fuel rod cross-sections (e.g. ⁇ a fuel rod absorption cross- section, ⁇ L f fuel rod fission energy release cross-section) based on these two parameters and make the fuel rod cross-sections match the real heterogeneity of the fuel assembly since it follows the history of each fuel rod through tracking the above two parameters.
- the fuel pin flux form factors correspond to the real assembly heterogeneity. Therefore, the method of this invention automatically captures the history of the fuel assembly and each individual fuel rod in time.
- the method of this invention doesn't need to perform different and complicated history calculations during fuel assembly data generation.
- the method of this invention follows the fuel pin true history over time and calculates the pin cell data (data over the incremental cross-sections) directly based on the fuel pins real history. Therefore, the method of this invention will be able to handle all kinds of control rod and discreet burnable absorber insertion and withdrawal scenarios.
- [0033J Unlike the full fuel pin by fuel pin calculations currently being studied in many national labs and universities, the method of this invention won't directly solve diffusion or transport equations for each pin (NGM-Next Generation Method). Instead, it adopts a 1.5-group-like method to simply adjust the pin-by-pin fluxes. Since there is no need for iteration and pin-by-pin coupling, this method is much faster than NGM while the transport results are well reproduced.
- the method of this invention requires very little computer processing unit time increase compared to the prior art method.
- the pin history data (burnup and fluence) are available in most design codes (for example, ANC). Therefore, there is no need to save any additional individual pin data. Individual fuel pin information is a large chunk of data. Saving any additional individual pin data will significantly increase the disk requirements and impact code performance which has been one of the biggest problems for NGM.
- the method of this invention will improve the prediction of pin power for any kind of history of control rod or discreet burnable absorber insertion or withdrawal.
- the method of this invention calculates individual cross-sections and fluxes that are needed for re-homogenization. That means that if this method is applied one is able to do re-homogenization in a very inexpensive and efficient way.
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- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- General Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- High Energy & Nuclear Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
Claims
Applications Claiming Priority (2)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
US2900508A | 2008-02-11 | 2008-02-11 | |
PCT/US2009/033160 WO2009134498A2 (en) | 2008-02-11 | 2009-02-05 | A methodology for modeling the fuel rod power distribution within a nuclear reactor core |
Publications (2)
Publication Number | Publication Date |
---|---|
EP2240939A2 true EP2240939A2 (en) | 2010-10-20 |
EP2240939A4 EP2240939A4 (en) | 2012-03-21 |
Family
ID=41255642
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
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EP09739300A Withdrawn EP2240939A4 (en) | 2008-02-11 | 2009-02-05 | A methodology for modeling the fuel rod power distribution within a nuclear reactor core |
Country Status (8)
Country | Link |
---|---|
EP (1) | EP2240939A4 (en) |
JP (1) | JP2011528101A (en) |
KR (1) | KR20100115754A (en) |
CN (1) | CN101946253A (en) |
BR (1) | BRPI0908354A2 (en) |
TW (1) | TW200949856A (en) |
WO (1) | WO2009134498A2 (en) |
ZA (1) | ZA201005627B (en) |
Families Citing this family (10)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
US8548789B2 (en) * | 2008-02-11 | 2013-10-01 | Westinghouse Electric Company Llc | Methodology for modeling the fuel rod power distribution within a nuclear reactor core |
FR2972839B1 (en) * | 2011-03-15 | 2013-03-29 | Areva Np | METHOD FOR OPTIMIZING THE PILOTAGE OF A PRESSURIZED WATER NUCLEAR REACTOR DURING LOAD MONITORING |
CN103617816B (en) * | 2013-10-29 | 2016-06-08 | 中国广核集团有限公司 | The measuring method of reactor core power distribution |
CN105404723B (en) * | 2015-10-30 | 2017-04-19 | 西安交通大学 | Method for precisely calculating power distribution of fuel assembly rod |
SI3510602T1 (en) * | 2016-09-06 | 2020-09-30 | Westinghouse Elecric Sweden Ab | A fuel assembly |
WO2019148420A1 (en) * | 2018-02-01 | 2019-08-08 | 国家电投集团科学技术研究院有限公司 | Reactor three-dimensional assembly information tracking method and system |
JP7349379B2 (en) * | 2020-01-28 | 2023-09-22 | 三菱重工業株式会社 | Fuel rod output analysis method, analysis device, and fuel rod output analysis program |
JP7438926B2 (en) * | 2020-12-22 | 2024-02-27 | 三菱重工業株式会社 | Replacement core design method and replacement core design system |
CN114840988B (en) * | 2022-04-22 | 2024-02-23 | 西安交通大学 | Automatic modeling method for pressurized water reactor core |
CN117807777B (en) * | 2023-12-27 | 2024-07-12 | 东北电力大学 | Test piece design method for realizing axial non-uniform heating of nuclear reactor fuel rod |
Citations (3)
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US4333797A (en) * | 1979-05-11 | 1982-06-08 | Hitachi, Ltd. | Reactor power control apparatus |
US20010026603A1 (en) * | 1998-03-17 | 2001-10-04 | Kabushiki Kaisha Toshiba | Reactor nuclear instrumentation system, reactor power distribution monitor system including above instrumentation system and reactor power distribution monitoring method |
US20060184286A1 (en) * | 2003-06-26 | 2006-08-17 | Framatome Anp Gmbh | Method for computer modeling the core of a nuclear reactor |
Family Cites Families (6)
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US4774050A (en) * | 1986-04-10 | 1988-09-27 | Westinghouse Electric Corp. | Axial power distribution monitor and display using outputs from ex-core detectors and thermocouples |
JP3023185B2 (en) * | 1991-01-17 | 2000-03-21 | 株式会社東芝 | Reactor core performance calculator |
JP3100069B2 (en) * | 1991-04-19 | 2000-10-16 | 株式会社日立製作所 | Reactor operation planning method and apparatus, and reactor core characteristic monitoring apparatus |
US20050135547A1 (en) * | 1998-03-11 | 2005-06-23 | Wolfgang Schulz | Control element for a nuclear reactor |
LU90570B1 (en) * | 2000-04-26 | 2001-10-29 | Europ Economic Community | Method of incineration of minor actinides in nuclear reactors |
JP3508021B2 (en) * | 2001-08-29 | 2004-03-22 | 株式会社原子力エンジニアリング | Reactor core calculation method |
-
2009
- 2009-02-05 BR BRPI0908354A patent/BRPI0908354A2/en not_active IP Right Cessation
- 2009-02-05 JP JP2010545982A patent/JP2011528101A/en active Pending
- 2009-02-05 WO PCT/US2009/033160 patent/WO2009134498A2/en active Application Filing
- 2009-02-05 KR KR1020107017784A patent/KR20100115754A/en not_active Application Discontinuation
- 2009-02-05 CN CN2009801048522A patent/CN101946253A/en active Pending
- 2009-02-05 EP EP09739300A patent/EP2240939A4/en not_active Withdrawn
- 2009-02-11 TW TW098104364A patent/TW200949856A/en unknown
-
2010
- 2010-08-05 ZA ZA2010/05627A patent/ZA201005627B/en unknown
Patent Citations (3)
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US4333797A (en) * | 1979-05-11 | 1982-06-08 | Hitachi, Ltd. | Reactor power control apparatus |
US20010026603A1 (en) * | 1998-03-17 | 2001-10-04 | Kabushiki Kaisha Toshiba | Reactor nuclear instrumentation system, reactor power distribution monitor system including above instrumentation system and reactor power distribution monitoring method |
US20060184286A1 (en) * | 2003-06-26 | 2006-08-17 | Framatome Anp Gmbh | Method for computer modeling the core of a nuclear reactor |
Non-Patent Citations (2)
Title |
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IWAMOTO AND M YAMAMOTO T: "Pin Power Reconstruction Methods of the Few-Group BWR Core Simulator NEREUS", JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, ATOMIC ENERGY SOCIETY OF JAPAN, JP, vol. 36, no. 12, 1 December 1999 (1999-12-01), pages 1141-1152, XP008139286, ISSN: 0022-3131 * |
See also references of WO2009134498A2 * |
Also Published As
Publication number | Publication date |
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TW200949856A (en) | 2009-12-01 |
BRPI0908354A2 (en) | 2019-09-24 |
WO2009134498A3 (en) | 2009-12-30 |
CN101946253A (en) | 2011-01-12 |
KR20100115754A (en) | 2010-10-28 |
EP2240939A4 (en) | 2012-03-21 |
JP2011528101A (en) | 2011-11-10 |
ZA201005627B (en) | 2011-06-29 |
WO2009134498A2 (en) | 2009-11-05 |
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