Patent Number: 048812471
Section: claims

1. A method of measuring the burnup of nuclear fuel comprising: (A) measuring the fast neutron counting rate of said nuclear fuel;  (B) reading said burnup off a curve which expresses the relationship between neutron emission rate and burnup for a nuclear fuel of comparable history, where the emission rate which corresponds to said neutron counting rate is obained by multiplying said neutron counting rate by the ratio of the neutron emission rate given by said curve for nuclear fuel of comparable history and known burnup to its similarly measured counting rate, and is defined by the formula EQU n/s=1.34.times.10.sup.-3 3.92  (A) measuring the fast neutron counting rate of said nuclear fuel;  (B) obtaining the burnup which corresponds to said counting rate from a fast neutron counting rate-burnup curve which is a neutron emission rate-burnup curve, where the scale of said counting rate given by said fast neutron counting rate-burnup curve is the product of (1) the scale of the emission rate of said neutron emission rate-burnup curve for nuclear fuel of comparable history and (2) the ratio of a similarly measured neutron counting rate of nuclear fuel of comparable history and known burnup to its neutron emission rate as given by said neutron emission rate-burnup curve, and is defined by the formula EQU n/s=1.34.times.10.sup.-3 3.92  where n/s equals neutron emission rate.  (A) means for measuring the fast neutron counting rate of said nuclear fuel;  (B) a curve giving the relationship between said fasat neutron counting rate and burnup which is defined by the formula EQU n/s=1.34.times.10.sup.-3 3.92  where n/s equals neutron emisson rate.  (A) means for measuring the fast neutron counting rate of nuclear fuel;  (B) a sample of nuclear fuel of comparable history and known burnup, for which, when placed in said means for measuring fast neutron counting rate, a fast neutron counting rate is obtainable;  (C) a neutron emission rate-burnup curve giving the relationship between neutron emission rate and burnup, from which the neutron emission rate of said sample is obtainable, and the ratio of said neutron emission rate to said neutron counting rate of said sample is determinable, so that when said nuclear fuel is placed in said means for measuring fast neutron counting rate and is fast neutron counting rate is measured, its fast neutron counting rate can be multiplied by said ratio to give its neutron emission rate and the burnup of said nuclear fuel can be read off said neutron emission rate-burnup curve, said curve being defined by the formula EQU n/s=1.34.times.10.sup.-3 3.92 2. A method according to claim 1 including the additional step of calculating said curve. 3. A method according to claim 1 wherein said fast neutron counting rate is measured with a boron-10 lined neutron detector or a U-235 lined fission detector which is shielded from gamma rays by a gamma ray shield and from thermal neutrons by a thermal neutron shield, where a moderator which slows fast neutrons down to thermal neutrons is provided inbetween said detector and said thermal neutron shield. 4. A method accordinig to claim 3 wherein said moderator is selected from the group consisting of water, polyethylene, and mixtures thereof. 5. A method according to claim 3 wherein said gamma rays are shielded with lead. 6. A method according to claim 3 wherein said thermal neutrons are shielded with cadmium. 7. A method of measuring the burnup of nuclear fuel comprising: 8. Apparatus for determining the burnup of nuclear fuel comprising: 9. Apparatus according to claim 8 which includes a boron-10 lined neutron detector or a U-235 lined fission detector to measure said fast neutron counting rate, where said detector is provided with a gamma ray shield and a thermal neutron shield, and a moderator which slows fast neutrons down to thermal neutrons is positioned inbetween said detector and said thermal neutron shield. 10. Apparatus accordiing to claim 8 wherein said moderator is selected from the group consisting of water, polyethylene, and mixtures thereof. 11. Apparatus according to claim 9 wherein said gamma ray shield is lead. 12. Apparatus according to claim 9 wherein said thermal neutron shield is cadmium. 13. Apparatus for measuring the burnup of nuclear fuel comprising: 14. Apparatus according to claim 13 which includes a boron-10 lined neutron detector or a U-235 lined fission detector to measure said fast neutron counting rate, where said detector is provided with a gamma ray shield and a thermal neutron shieldd, and a moderator which slows fast neutrons down to thermal neutrons is positioned in between said detector and said thermal neutron shield. 15. Apparatus according to claim 14 wherein said moderator is selected from the group consisting of water, polyethylene, and mixtures thereof. 16. Apparatus according to claim 14 wherein said gamma ray shield is lead. 17. Apparatus according to claim 14 where said thermal neutron shield is cadmium. 18. Apparatus according to claim 8 wherein the scale of said fast neutron counting rate given by said curve is the product of the scale of the neutron emission rate of a neutron emission rate-burnup curve times the ratio of a similarly measured fast neutron couning rate of nuclear fuel of known burnup and comparable history to its neutron emission rate as given by said neutron emission rate-burnup curve.