Search

Performance tests of mechanical decladding technology for estimating the feeding portions of the recovered fuel fragments to an electrolytic reduction process were conducted in terms of the fuel rod burnups of 27.3 to 65.7 GWd/tonne uranium (tU) for the used pressurized water reactor nuclear fuel. The decladding efficiencies with fuel burnups were quantitatively obtained from slitting decladding tests. Based on the average fuel rod burnups, fuel rods with an average burnup of up to 52.3 GWd/tU showed above 99%, but higher burnup fuels of above 54.9 GWd/tU were below 97.52% in the decladding efficiency. It was interpreted that variations in decladding efficiency with fuel burnups were closely linked to the opening characteristics of the gap between the pellets and cladding. However, the fuel fragment size distribution after slitting decladding has little difference in fuel burnup changes between 34.8 and 55.4 GWd/tU. Hence, feeding portions of the fuel fragments from an assembly basis by using the decladding efficiency and recovered fragment size distribution data were estimated with burnup variations of 35 to 52.5 GWd/tU. [ABSTRACT FROM AUTHOR]

In closed chamber distillation equipment, the distillation characteristics of LiCl-KCl eutectic salt containing rare earth phosphates were investigated to establish an operating condition to maximize the operating efficiency o f the equipment. To understand the behaviors of LiCl-KCl eutectic salt in the distillation chamber, changes in the pressure in the distillation equipment were observed with a change in the temperature gradients between the vaporization chamber and condensation chamber, and the temperature distributions in the chamber were modeled using computational fluid dynamics software (ANSYS CFX). Based on these results, a model of the salt behavior in the chamber was established, and the most effective operating condition for over 99 wt% of salt recovery was determined. Many distillation experiments of LiCl-KCl eutectic salt containing rare earth phosphates were petformed at the condition, and it was confirmed that the salt recovery efficiency (>99 wt%) could be consistently achieved at the condition. [ABSTRACT FROM AUTHOR]

As an alternative to conventional Group I and II separation methods (such as adding a chemical agent and ion exchange), melt crystallization processes, zone freezing, and layer melt crystallization were tested for the separation (or concentration) of cesium and strontium fission products in a LiCl waste salt generated from an electrolytic reduction process of a spent oxide fuel. in these melt crystallization processes, impurities (CsCl and SrCl2) are concentrated in a small fraction of the LiCl salt by the solubility difference between the melt phase and the crystal phase. As experimental variables, initial molten salt temperature, crucible rising velocity in the zone freezing case, and cooling airflow rate in the layer crystallization case were used. In the zone freezing process, although the operating time is long (1.7 mm/h of crucible rising velocity) when assuming a LiCl salt reuse rate of 90 wt%, >90% separation efficiency for both CsCl and SrCl2 was shown. In the layer crystallization process, the crystal growth rate strongly affects the crystal structure and therefore the separation efficiency. At a 25 to 30 e/min cooling airflow rate, 700 to 710°C initial molten salt temperature, and <5 g/min crystal growth rate, the separation efficiency of both CsCl and SrCl2 exceeded 90% by the layer crystallization process, assuming a LiCl salt reuse rate of 90 wt%. [ABSTRACT FROM AUTHOR]

A fundamental study on the distillation rate on LiCl-KCl eutectic salt under different vacuums from 66 to 6600 Pa (0.5 to 50 mm Hg) was performed by using both a non isothermal and an isothermal thermogravimetric (TG) analysis. Based on the nonisothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A salt distillation operation with a moderated distillation rate of 10-5 to 10-5 mol·cm-2·s-1 is possible at temperatures of <1300 K and vacuums of 660 to 6600 Pa. An -99% salt distillation efficiency was obtained after 1 h at a temperature above 1150 K under 6600 Pa. An increase in the vaporizing surface area is relatively effective for removing residual salt in the remaining particles, when compared to that for the vaporizing time. More than 99.95% of total distillation efficiency was obtained for a 1-h distillation operation by increasing the inner surface area from 4.52 to 12.56 cm² (about 3 times increase). [ABSTRACT FROM AUTHOR]

In this work, a new approach to remove fission products including decay heat elements was proposed. This study aims at providing a new way to minimize the amount of waste salt for a repository, while removing the high decay heat fission products [Cs, Sr, Ba, and Y including other rare earth (RE) elements] from the waste salts generated during a chloride pyroprocessing procedure. These elements were removed in consecutive order from the pyroprocessing units. First, Cs could be released in the form of an oxide gas during voloxidation of UO2 and captured by a fly-ash filter. Then, Sr was recovered in the form of carbonate precipitates from the LiCl waste salt generated during the course of an electoreduction process, by using Li2CO3. Finally, RE elements plus yttrium in the spent LiCl-KCl waste salt generated during electrorefining were removed in the form of oxides (or oxychlorides) by using an oxygen sparging method. It was confirmed that the removal yields of each element were ~90% for Cs at ~1473 K, >99% for Sr at a molar ratio of [Li2CO3/SrCl2 = 3], and >99% for the RE elements plus yttrium. Using these successes as a basis, a reference flow sheet for removing the high decay heat elements from pyroprocessing units is presented in this work. Also, a salt regeneration system to minimize the amount of waste salt is proposed in this study. [ABSTRACT FROM AUTHOR]