Patent Number: 
Section: claims

1. A method for determining respective amounts of materials in a first cell, said first cell including a plurality of principal materials, the method comprising:(a) selecting a neighbor to a principal material in said first cell, wherein said selected neighbor is one of a plurality of agent materials;(b) determining a proxy amount of said selected neighbor for said first cell, wherein said proxy amount represents an amount of said selected neighbor that approximates a macroscopic property of said principal material as a function of an amount of said principal material in said first cell;(c) performing acts (a) and (b) for each principal material of the plurality of principal materials in said first cell; and(d) determining a summed proxy amount of each agent material for said first cell by summing agent material proxy amounts associated with said each agent material. 2. The method of claim 1, wherein an agent material of the plurality of agent materials is an isotope having a plurality of acceptably characterized microscopic properties of interest. 3. The method of claim 1, wherein an agent material of the plurality of agent materials is a fictional material having fictional values for a plurality of microscopic properties of interest. 4. The method of claim 1, wherein a principal material is an isotope having at least one microscopic property that is not acceptably characterized. 5. The method of claim 1, wherein a number of acceptably characterized microscopic properties of at least one agent material is larger than a number of acceptably characterized microscopic properties of at least one principal material. 6. The method of claim 1, wherein agent materials and principal materials include fission products. 7. The method of claim 1, wherein a first cell corresponds to a region of a plurality of regions of a nuclear reactor system. 8. The method of claim 1, wherein a first cell corresponds to a region of a plurality of regions of a simulated nuclear reactor system. 9. The method of claim 1, wherein a first cell corresponds to a region of a plurality of regions of an existing nuclear reactor system. 10. The method of claim 1, wherein a first cell corresponds to a region of a plurality of regions of a currently operating nuclear reactor system. 11. The method of claim 1, wherein said first cell corresponds to region having a homogeneous composition of materials. 12. The method of claim 1, wherein said first cell corresponds to a region including a heterogeneous composition of materials. 13. The method of claim 1, further comprising performing acts (a)-(d) for each cell of a plurality of cells. 14. The method of claim 13, wherein said first cell corresponds to a region having a first shape different than a second shape of a second region corresponding to a second cell of said plurality of cells. 15. The method of claim 13, wherein said first cell corresponds to a first region having a first volume different than a second volume of a second region corresponding to a second cell of said plurality of cells. 16. The method of claim 1, wherein act (d) comprises determining summed proxy amounts for a plurality of agent materials. 17. The method of claim 1, wherein act (a) comprises selecting said neighbor from among a plurality of potential neighbors based on a comparison of a microscopic property of said principal material with a microscopic property of each of said plurality of neighbors. 18. The method of claim 17, further comprising identifying said plurality of potential neighbors from among said agent materials based on a comparison of an atomic mass number (A) of said principal material with an A of each of said agent materials. 19. The method of claim 17, wherein said microscopic property is a microscopic absorption cross section. 20. The method of claim 19, further comprising approximating said microscopic absorption cross section with a single value based on a neutron flux spectrum. 21. The method of claim 19, further comprising approximating said microscopic absorption cross section based on a function of a neutron flux spectrum. 22. The method of claim 19, further comprising approximating said microscopic absorption cross section based on an integral of a function of a microscopic absorption cross section weighted by a neutron flux spectrum. 23. The method of claim 17, wherein said microscopic property is a microscopic scattering cross section. 24. The method of claim 23, further comprising approximating said microscopic scattering cross section with a single value based on a neutron flux spectrum. 25. The method of claim 23, further comprising approximating said microscopic scattering cross section based on a function of a neutron flux spectrum. 26. The method of claim 1, wherein said macroscopic property is a macroscopic absorption cross section. 27. The method of claim 1, wherein said macroscopic property is a macroscopic scattering cross section. 28. The method of claim 1, wherein act (a) comprises selecting an agent material from among a plurality of fission products. 29. The method of claim 1, wherein act (a) comprises selecting an agent material from among a plurality of fission products in a peak of a fission product yield curve containing said principal material. 30. The method of claim 1, wherein a potential neighbor having a microscopic property of approximately zero is ignored. 31. The method of claim 1, wherein act (d) comprises adding an amount of said each agent material to a sum of proxy amounts associated with said each agent material. 32. The method of claim 1, further comprising:(e) determining a flux in said first cell based on said summed proxy amounts of said each agent material in said first cell. 33. The method of claim 32, wherein said determining of act (e) is further based on one or more summed proxy amounts of one or more agent materials in a second cell. 34. The method of claim 32, wherein said determining of act (e) is further based on one or more summed proxy amounts of one or more agent materials in a plurality of cells. 35. The method of claim 32, further comprising determining a neutron flux spectrum. 36. The method of claim 32, wherein act (e) comprises determining said flux using a Monte Carlo method. 37. The method of claim 1, further comprising:(e) determining an updated amount of one of said materials based on an amount of said one of said materials and a flux in said first cell. 38. The method of claim 1, further comprising:moving a neutron absorber in a nuclear reactor system. 39. The method of claim 1, further comprising:automatically moving a neutron absorber in a nuclear reactor system.