METHOD OF EVALUATING NEUTRON EMISSION RATE FROM SPENT FUEL, AND PROGRAM THEREFOR

摘要

PROBLEM TO BE SOLVED: To evaluate a neutron emission rate of Cm244 from a spent fuel. SOLUTION: A generation amount of a neutron emission nuclide Cm244 provided by combustion calculation based on respective data when loaded in a core of a nuclear reactor to be burnt by neutron irradiation is conformed substantially with an actual generation amount by regulating a neutron capturing cross-sectional area(s) of Am243 or Pu242, or Am243 and Pu242 using combustion calculation combined with a neutron spectrum calculation code 2 and a nuclear data library 3, based on a specification 1 of a fuel assembly as an evaluation object. A substantial consistency with an actually generated neutron emission rate (measured value) E is allowed thereby without being affected substantially by a neutron multiplying characteristic, and calculation precision 8 of the neutron emission rate based on Cm244 is enhanced thereby to conduct the neutron emission rate evaluation 9 of Cm244.