Comparison of spent fuel data bases for K basin

The SCATS and reconstructed SCATS databases representing spent fuel in the Hanford K basins are compared

Review of statistical analysis of trapped gas

A review was conducted of trapped gas estimates in Hanford waste tanks. Tank waste levels were found to correlate with barometric pressure changes giving the possibility to infer amounts of trapped gas. Previous models of the tank waste level were extended to include other phenomena such as evaporation in a more complete description of tank level changes

MCNP criticality validation and bias for LEU systems

The bias in MCP calculations was evaluated for low enriched uranium (LEU) systems typical of N reactor fuel. A formula that includes the bias and its uncertainties is given to ensure that LEU systems are safely subcritical

DKPRO: A radionuclide decay and reprocessing code

The DKPRO code solves the general problem of modeling complex nuclear wastes streams using ORIGEN2 radionuclide production files. There is a continuing need for estimates of Hanford radionuclides. Physical measurements are one basis; calculational estimates, the approach represented here, are another. Given a known nuclear fuel history, it is relatively straightforward to calculate radionuclide inventories with codes such as the widely-used Oak Ridge National Laboratory code ORIGEN2

Consequence ranking of radionuclides in Hanford tank waste

Radionuclides in the Hanford tank waste are ranked relative to their consequences for the Low-Level Tank Waste program. The ranking identifies key radionuclides where further study is merited. In addition to potential consequences for intrude and drinking-water scenarios supporting low-level waste activities, a ranking based on shielding criteria is provided. The radionuclide production inventories are based on a new and independent ORIGEN2 calculation representing the operation of all Hanford single-pass reactors and the N Reactor

ORIGEN2 calculations supporting TRIGA irradiated fuel data package

ORIGEN2 calculations were performed for TRIGA spent fuel elements from the Hanford Neutron Radiography Facility. The calculations support storage and disposal and results include mass, activity,and decay heat. Comparisons with underwater dose-rate measurements were used to confirm and adjust the calculations

Decay heat for the Fast Test Reactor (FTR)

The decay heat for the Fast Test Reactor (FTR) is reevaluated for cooling times up to 30 years. Decay heat from both fission-products and actinides is included. Fission-product decay data used in this evaluation are taken from ENDF/B version IV. Consideration is given to the effects of fuel cycles and loadings and to the importance of fissions in the minor fuels such as /sup 235/U, /sup 240/Pu, and /sup 241/Pu, as well as in the major fuels, /sup 238/U and /sup 239/Pu. Decay-heat uncertainties are evaluated and are typically less than 5 percent for cooling times beyond 50 seconds. The effective energy release per fission for FTR is also reevaluated