[en] The Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program has the objective of developing a cost-competitive enrichment process that will ultimately replace the gaseous diffusion process used in the United States. Current nuclear fuel fabricators are set up to process only the UF₆ product from gaseous diffusion enrichment. Enriched uranium-iron alloy from the U-AVLIS separator system must be chemically converted into an oxide form acceptable to these fabricators to make fuel pellets that meet American Society for Testing and Materials (ASTM) and utility company specifications. A critical step in this conversion is the modified direct denitration (MDD) that has been selected and presented in the AVLIS Conceptual Design for converting purified uranyl nitrate to UO₃ to be shipped to fabricators for making UO₂ pellets for power reactor fuel. This report describes the MDD process, the equipment used, and the experimental work done to demonstrate the conversion of AVLIS product to ceramic-grade UO₃ suitable for making reactor-grade fuel pellets

[en] Review of the Three Mile Island accident by NRC has resulted in new post-accident-sampling-capability requirements for utilities that operate pressurized water reactors and/or boiling water reactors. Several vendors are offering equipment that they hope will suffice to met both the new NRC regulations and an operational deadline of January 1, 1981. The advantages and disadvantages of these systems and projected future-new-system needs for TVA reactors are being evaluated in light of TMI experience

[en] Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with ²³⁸U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial

[en] A cost-benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model nuclear fuel reprocessing plant which processes light-water reactor (LWR) fuels, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term ''as low as practicable'' in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment systems are added to the base case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitments are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs and the radiological doses, detailed calculations, and tabulations is presented in Appendix A and ORNL-4992. (U.S.)

No abstract available

[en] The principal objective of ORNL-ESP is to demonstrate process monitoring as it might be accomplished by inspectors of any nuclear fuel recycle facility. Improved instrumentation and computer interfacing, currently being installed, provide the ORNL ²³³U Pilot Plant with the capability of a dynamic volume balance in the solvent extraction system. Later, an accurate, (almost) instantaneous fissile mass balance will be routinely obtainable in the Pilot Plant. Subsidiary objectives include minimizing MUF/LEMUF, detecting material diversions, and alerting appropriate authorities in control of the facility in case of process anomalies. A continuing program will examine technology which might be utilized for facility design. Ultimately, process monitoring/control integrated with safeguards can convert the ORNL ²³³U Pilot Plant into a partial safeguards demonstration facility

[en] Additional results confirm that during most of the consolidation of polycrystalline salt in brine, the previously proposed rate expression applies. The final consolidation, however, proceeds at a lower rate than predicted. The presence of clay hastens the consolidation process but does not greatly affect the previously observed relationship between permeability and void fraction. Studies of the migration of brine within polycrystalline salt specimens under stress indicate that the principal effect is the exclusion of brine as a result of consolidation, a process that evidently can proceed to completion. No clear effect of a temperature gradient could be identified. A previously reported linear increase with time of the reciprocal permeability of salt-crystal interfaces to brine was confirmed, though the rate of increase appears more nearly proportional to the product of sigma ΔP rather than sigma ΔP² (sigma is the uniaxial stress normal to the interface and ΔP is the hydraulic pressure drop). The new results suggest that a limiting permeability may be reached. A model for the permeability of salt-crystal interfaces to brine is developed that is reasonably consistent with the present results and may be used to predict the permeability of bedded salt. More measurements are needed, however, to choose between two limiting forms of the model

[en] The principal objective of the ORNL Integrated Safeguards Program (ISP) is to provide enhanced material accountability, improved process control, and greater security for nuclear fuel cycle facilities. With the improved instrumentation and computer interfacing currently installed, the ORNL ²³³U Pilot Plant has demonstrated capability of a near-real-time liquid-volume balance in both the solvent-extraction and ion-exchange systems. Future developments should include the near-real-time mass balancing of special nuclear materials as both a static, in-tank summation and a dynamic, in-line determination. In addition, the aspects of site security and physical protection can be incorporated into the computer monitoring

[en] The principal objective of the ORNL Integrated Safeguards Program (ISP) is to provide enhanced material accountability, improved process control, and greater security for nuclear fuel cycle facilities. With the improved instrumentation and computer interfacing currently installed, the ORNL ²³³U Pilot Plant has demonstrated capability of a near-real-time liquid-volume balance in both the solvent-extraction and ion-exchange systems. Future developments should include the near-real-time mass balancing of special nuclear materials as both a static, in-tank summation and a dynamic, in-line determination. In addition, the aspects of site security and physical protection can be incorporated into the computer monitoring

[en] A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials from a model nuclear fuel reprocessing plant which processes light-water reactor (LWR) fuels, and to determine the radiological impact (dose commitment) of the released materials on the environment. The study is designed to assist in defining the term as low as reasonably achievable in relation to limiting the release of radioactive materials from nuclear facilities. The base case model plant is representative of current plant technology and has an annual capacity of 1500 metric tons of LWR fuel. Additional radwaste treatment systems are added to the base case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The cost for the added waste treatment operations and the corresponding dose commitments are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases is in an early stage of development and is not suitable for immediate use. The methodology used in estimating the costs, and the radiological doses, detailed calculations, and tabulations are presented in Appendix A and ORNL-4992. This report is a revision of the original study