Abstract
An accurate treatment of neutron transport and depletion in modern fuel assemblies, which are characterized by heterogeneous, complex designs, and higher fuel enrichments and burnups, requires the use of advanced and complex computational tools capable of simulating multidimensional geometries. Most of the modern computational systems developed for this purpose are built through coupling of multidimensional neutron transport and point depletion codes. For many routine applications, the use of these types of tools may not be efficient, as they may require large computational time, significant computer resources, additional specialized knowledge on the methods used by the code or more extensive training of the code user. In these cases, depending on the specific type of application and the level of accuracy required, the user may benefit from the availability of faster and easier to use alternative tools. The ORIGEN-ARP code in SCALE [1] could serve as such an alternative tool for depletion, decay, and source term analyses.
Validation and verification studies of ORIGEN-ARP for various applications have shown that it provides results with an accuracy level comparable to that provided by more complex codes. As in other computational tools, the accuracy level in ORIGEN-ARP is greatly influenced by the adequacy and quality of the data in the crosssection libraries used. The latest release of SCALE includes ORIGEN-ARP cross section libraries generated for a wide range of modern assembly designs that are currently used in the nuclear industry [2]. Results of validation studies for the libraries developed for VVER assembly designs, as well as for more recently generated libraries for RBMK assembly configurations, are presented and discussed in this paper. The validation was performed against isotopic assay measurement data for spent fuel.
