An Investigation Comparing the Criticality of Stored DOE Waste Using MCNP with Previously Published Results Obtained with KENO
D-Ok. Choe, B. N. Shelkey, D. M. Slaughter
Neutron transport codes have been developed to solve a variety of difficult issues that include the transport of radionuclides from known sources, the criticality of sources contained and surrounded in complex geometries, and the flux characteristics of nuclear reactors. The code KENO V.a was previously used by Lewis to determine criticality of various waste configurations in storage drums containing up to 2.0 %wt U-235 in an UF4 and oil matrix at the Paducah Gaseous Diffusion Plant in Kentucky. KENO predicted values of keff+2s below a predetermined subcritical maximum value of 0.9609 for all waste configurations. In the current study, the waste configurations were duplicated and modeled using the code MCNP-4B2. The results show criticality greater than 1.0 for all waste configurations except one. A primary reason for this divergence of values between the two codes involves the use of a continuous energy function in MCNP versus a non-continuous energy function in KENO. A greater sophistication of geometry specifications is possible with MCNP. Unlike MCNP, there are geometry restrictions inherent in the KENO code which inhibits the functionality of the code, unlike MCNP. The use of the KENO code may result in an underestimation of possible radiation exposure due to its more simplified format. Generating accurate simulations for estimating radiation fields have significant implication on creating an effective as low as reasonably achievable (ALARA) program.
This abstract was presented at the 34th Annual Midyear Meeting, "Radiation Safety and ALARA Considerations for the 21st Century", Automated Applications Session, 2/4/2001 - 2/7/2001, held in Anaheim, CA.