Management of intermediate-level radioactive wastes in the United States

While used extensively, the term intermediate-level waste is not a clearly defined waste category. Assuming the ILW includes all radioactive wastes requiring shielding but not ordinarily included in a high-level waste canister, its major sources include power plant operations, spent fuel storage, and spent fuel reprocessing. While the volume is approx. 10/sup 2/ greater than that of high-level waste, ILW contains only approx. 1% of the radioactivity. Power plant waste, constituting approx. 87% of the waste volume, is generally nontransuranic waste. The other approximately 13% from fuel reprocessing is generally transuranic. Intermediate-level wastes fall into the general categories of highly radioactive hardware, failed equipment, HEPA filters, wet wastes, and noncombustible solids. Within each category, however, the waste characteristics can vary widely, necessitating different treatments. The wet wastes, primarily power plant resins and sludges, contribute the largest volume; fuel hulls and core hardware represent the greatest activity. Numerous treatments for intermediate-level wastes are available and have been used successfully. Packaging and transportation systems are also available. Intermediate-level wastes from power plants are disposed of by shallow-land burial. However, the alpha-bearing wastes are being stored pending eventual disposal to a geologic repository or by other means, e.g., intermediate-depth burial, sea disposal. Problem areas associated with intermediate-level wastes include: disposal criteria need to be established; fixation of organic ion exchange resins from power plant operation needs improvement; and reprocessing of LWR fuels will produce ILW considerably different from power plant ILW and requiring different treatment

ISOSHLD 4.0

ISOSHLD was developed to carry out routine shielding calculations that would otherwise be very laborious. More sophisticated methods exist, but there is still need for easy-to-use reliable computer codes. The code has been revised several times including PC and mainframe-based versions. These versions generally represent different objectives and differ in code, data, and supported features. The current version includes all features. It has a 30 group energy range from 10 keV and 10 MeV. Bremsstrahlung is included, MODE 1 capability is retained (fission product inventories from decay and transmutation), and the libraries now fully support the actinides. There were several design objectives in this latest revision; modernization, state-of-the-art nuclear data, and full capability including source generation and time dependent calculations

85Kr management trade-offs: a perspective to total radiation dose commitment

Radiological consequences arising from the trade-offs for /sup 85/Kr waste management from possible nuclear fuel resource recovery activities have been investigated. The reference management technique is to release all the waste gas to the atmosphere where it is diluted and dispersed. A potential alternative is to collect, concentrate, package and submit the gas to long-term storage. This study compares the radiation dose commitment to the public and to the occupationally exposed work force from these alternatives. The results indicate that it makes little difference to the magnitude of the world population dose whether /sup 85/Kr is captured and stored or chronically released to the environment. Further, comparisons of radiation exposures (for the purpose of estimating health effects) at very low dose rates to very large populations with exposures to a small number of occupationally exposed workers who each receive much higher dose rates may be misleading. Finally, cost studies (EPA 1976 and DOE 1979a) show that inordinate amounts of money will be required to lower this already extremely small 80-year cumulative world population dose of 0.05 mrem/person (<0.001% of natural background radiation for the same time period)

Operational dose rate visualization techniques

The analysis of the gamma ray dose rate in the vicinity of a radiation source can be greatly aided by the use of recent state-of-the-art visualization techniques. The method involves calculating dose rates at thousands of locations within a complex geometry system. This information is then processed to create contour plots of the dose rate. Additionally, when these contour plots are created, animations can be created that dynamically display the dose rate as the shields or sources are moved

Application of ALARA principles to shipment of spent nuclear fuel

The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose

Waste isolation safety assessment program. Technical progress report for FY-77

Purpose of WISAP is to evaluate the post-closure effectiveness of deep geologic nuclear waste repository systems. The work conducted centered in four subject areas: (1) the analysis of potential repository release scenarios, (2) the analysis of potential release consequences, (3) the measurement of waste form leaching rates, and (4) the measurement of the interactions of dissolved radionuclides with geologic media. 12 figures, 24 tables