3 research outputs found
Recommended from our members
Neutron and photon effective dose equivalent rate calculations for the repackaging of tru waste
Neutron and photon effective dose equivalent rates were estimated for operations that will occur in the characterization and repackaging of transuranic (TRU) waste drums. These activities will be performed in structures called Mobile Units (MU). A MU is defined as a modular and transportable container, also called a transportainer. The transportainers have been designed to house a process required for certification of TRU wastes. The purpose of these calculations was to provide dose rates from Pu-238 TRU waste in various locations in the transportainer using MCNP-4C. In addition to dose rates for the various radiological operations in the repackaging area, the dose rate from the adjacent storage area was calculated to determine the contribution to the total dose rate
Recommended from our members
Continued development of the next generation of SOURCES
In development for over two decades, the SOURCES computer code calculates neutron production rates and spectra from four types of problems: homogeneous media, two-region interfaces, three-region interfaces and that of a monoenergetic alpha particle beam incident on a slab of target material. In terms of file structure, SOURCES consists of a usercreated input file (tapel), several data files (tapes 2-5) and up to six output files (outp, outp2, and tapes 6-9). SOURCES was recently upgraded from version 4A to version 48 when the code's decay data library, tapeti, was updated for 44 of the 105 available decay sources. The new version provided spontaneous fission information for 252Cf and alpha decay data revisions for 43 isotopes and isomers having Watt fission spectra parameters.' Following that work, this summary documents similar modifications made to the remaining 61 sources. It also summarizes additional enhancements and planned improvements to SOURCES for the next major code release
Recommended from our members
Extremity model for neutron dose calculations
In personnel dosimetry for external radiation exposures, health physicists tend to focus on measurement of whole body dose, where 'whole body' is generally regarded as the torso on which the dosimeter is placed.' Although a variety of scenarios exist in which workers must handle radioactive materials, whole body dose estimates may not be appropriate when assessing dose, particularly to the extremities. For example, consider sources used for instrument calibration. If such sources are in a contact geometry (e.g. held by fingers), an extremity dose estimate may be more relevant than a whole body dose. However, because questions arise regarding how that dose should be calculated, a detailed extremity model was constructed with the MCNP-4Ca Monte Carlo code. Although initially intended for use with gamma sources, recent work by Shores2 provided the impetus to test the model with neutrons