Detailed Design of ^242mAm Breeding in Pressurized Water Reactors

There is growing interest in the use of ^242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of ^242mAm is very high, the best way to obtain ^242mAm is by the capture of fast or epithermal neutrons in ²⁴¹Am. However, fast spectrum reactors are not readily available.

In this paper, we explore the possibility of producing ^242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the ^242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes ^242mAm destruction. Since ^242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed.

In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of ^242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW(electric)yr] of ^242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ˜1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the ^242mAm breeding targets were evaluated and found to be close to those of a standard PWR.

Another product of neutron capture in the ²⁴¹Am reaction is ²³⁸Pu. It was found that in a typical 1000 MW(electric) PWR core with one-third of the fuel assemblies containing ²⁴¹Am targets, up to 15.1 kg of ²³⁸Pu enriched to 80% can be produced per year.