Method for the destruction of plutonium by irradiation in a light water reactor

#This invention concerns the destruction of plutonium, especially of weapons grade plutonium.

#The relaxation in tension between the United States and the former Soviet Union, witnessed over the past few years, has given rise to tens of thousand of nuclear warheads on both sides which have been declared excess to current needs. According to a recently published U.S. National Academy of Sciences report on Management and Disposition of Excess Weapons Plutonium, the excess amounts to 50 tons of weapons grade plutonium (W-Pu) in the U.S.A. At a recent NATO Advanced Research Workshop on managing the plutonium surplus Russian experts consider the Russian excess of W-Pu to be 100 tons.

#At present W-Pu and highly enriched uranium (W-U) are present in warheads and will have to be transferred for dismantling. In this process a major problem will arise with regard to proliferation of nuclear weapons to national powers or sub-national groups by theft or unauthorised diversion.

#The invention will now be described in more detail by means of a preferred embodiment and the annexed drawings.

#FIG. 1 shows the main pathways for the production of .sup.238 Pu from .sup.239 Pu and .sup.235 U.

#FIG. 2 shows the Pu isotopic masses per ton of matrix: in fresh fuel (A) and after 1, 2, and 2.3 cycles (B, C, D respectively) in a PWR (one cycle=878 days). Fresh fuel contains 2.5% W-Pu+2.5% W-U in an inert matrix.

#The burnup calculations are based upon a standard pressurised water reactor with 3.2% .sup.235 U initial fuel enrichment, power level of 37.5 MW/ton, and burnup of 33 GWd/ton (achieved after 879 EFPD-effective full power days). To simulate the irradiation of a pin with a different fuel in this reactor, the following procedure was adopted: An initial calculation was made for the standard fuel configuration (i.e. 3.2% .sup.235 U, 33 GWd/t) to determine the average neutron flux during irradiation.

#Using this average neutron flux and the same one group cross sections (thereby ensuring that the neutron flux and spectrum correspond to a 3.2% .sup.235 U enriched fuel with a burnup of 33 GWd/t over 879 EFPD), a new fuel configuration can be defined e.g. inert matrix plus fissile material or as MOX, and the calculations are then repeated for the irradiation period of 879 EFPD.

#A difficulty which arises with the above procedure is that in the initial calculation burnup dependent cross sections are used to account for changes in the neutron energy spectrum brought about by the buildup of additional fissionable isotopes. Depending on the burnup, different cross sections are used especially for the plutonium isotopes. If a different fuel type such as the inert matrix-based fuel is used, the code assumes the core consists entirely of this material. The burnup of this material in the previously defined neutron energy spectrum is clearly different from the standard fuel but it is the inert matrix-based fuel which determines the burnup and therefore which cross sections are to be used.