A Description of the Ceramic Waste Form Production Process from the Demonstration Phase of the Electrometallurgical Treatment of EBR-II Spent Fuel

2001 ◽  
Vol 134 (3) ◽  
pp. 263-277 ◽  
Author(s):  
Michael F. Simpson ◽  
K. Michael Goff ◽  
Stephen G. Johnson ◽  
Kenneth J. Bateman ◽  
Terry J. Battisti ◽  
...  
Author(s):  
Kenneth J. Bateman ◽  
Charles W. Solbrig

The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste form is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm in length during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas through the center hole, also works well as long as the product of heat capacity and velocity of the gas is equivalent to that of the flowing aluminum, and the velocity is high enough to produce an intermediate size heat transfer coefficient. The fourth method, using an electric heater, works well and heater sizes between 500 to 1000 Watts are adequate. These later three methods all can reduce the heatup time to 44 hours allowing production to be doubled and a more uniform heating.


2017 ◽  
Vol 4 ◽  
Author(s):  
Eric R. Vance ◽  
Dorji T. Chavara ◽  
Daniel J. Gregg

ABSTRACTSynroc has evolved over the last 40 years from the titanate full-ceramics developed in the late 1970s to a technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages in terms of waste loading and suppressing volatile losses.A first of a kind Synroc plant for immobilizing intermediate level waste arising from Mo-99 production is currently in detailed engineering at ANSTO.Since the year 2000, Synroc has evolved from the titanate full-ceramics developed in the late 1970s to a technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages in terms of waste loading and suppressing volatile losses. Furthermore recent efforts have focused strongly on waste form development for plutonium-bearing wastes in the UK, for different options for the immobilization of Idaho calcines and most recently developing an engineered waste form for the intermediate level wastes arising from 99Mo production, for the Australian Nuclear Science and Technology Organisation (ANSTO). A variety of other studies are currently in progress, including engineered waste forms for spent fuel and investigating the proliferation risks for titanate-based waste forms containing highly enriched uranium or plutonium. This paper also attempts to give some perspective on Synroc waste forms and process technology development in the nuclear waste management industry.


2002 ◽  
Vol 713 ◽  
Author(s):  
Roman V. Bogdanov ◽  
Yuri F. Batrakov ◽  
Elena V. Puchkova ◽  
Andrey S. Sergeev ◽  
Boris E. Burakov

ABSTRACTAt present, crystalline ceramic based on titanate pyrochlore, (Ca,Gd,Hf,Pu,U)2Ti2O7, is considered as the US candidate waste form for the immobilization of weapons grade plutonium. Naturally occuring U-bearing minerals with pyrochlore-type structure: hatchettolite, betafite, and ellsworthite, were studied in orders to understand long-term radiation damage effects in Pu ceramic waste forms. Chemical shifts (δ) of U(Lδ1)– and U(Lβ1) – X-ray emission lines were measured by X-ray spectrometry. Calculations were performed on the basis of a two-dimensional δLá1- and δLδ1- correlation diagram. It was shown that 100% of uranium in hatchettolite and, probably, 95-100% of uranium in betafite are in the form of (UO2)2+. formal calculation shows that in ellsworthite only 20% of uranium is in the form of U4+ and 80% of the rest is in the forms of U5+ and U6+. The conversion of the initial U4+ ion originally occurring in the pyrochlore structure of natural minerals to (UO2)2+ due to metamict decay causes a significant increase in uranium mobility.


2013 ◽  
Vol 1518 ◽  
pp. 73-78 ◽  
Author(s):  
Shirley K. Fong ◽  
Brian L. Metcalfe ◽  
Randall D. Scheele ◽  
Denis M. Strachan

ABSTRACTA calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.


MRS Advances ◽  
2018 ◽  
Vol 3 (20) ◽  
pp. 1059-1064 ◽  
Author(s):  
Eric R. Vance ◽  
Dorji T. Chavara ◽  
Daniel J. Gregg

Abstract:Since the year 2000, Synroc has evolved from the titanate full-ceramic waste forms developed in the late 1970s to a hot isostatic pressing (HIP) technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages over vitrification in terms of, for example, waste loading and suppressing volatile losses. This paper describes recent progress on waste form development for intermediate-level wastes from 99Mo production at ANSTO, spent nuclear fuel, fluoride pyroprocessing wastes and 129I. The microstructures and aqueous dissolution results are presented where applicable. This paper provides perspective on Synroc waste forms and recent process technology development in the nuclear waste management industry.


1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.


1983 ◽  
Vol 26 ◽  
Author(s):  
Walter J. GRAY ◽  
Gary L. McVay ◽  
John. O. Barner ◽  
John W. Shade ◽  
Roger W. Cote

ABSTRACTLeach tests have been performed on spent fuel in synthetic Permian Basin salt brine at 25 and 75°C. Complementary tests on unirradiated UO2 pellets have been conducted in both salt brine and deionized water in the range 25 to 150°C. Iron and/or oxidized zircaloy coupons were included in some of the tests. Uranium release from spent fuel was more than 100 times greater than from U02. In brine, iron had no significant effect on the total uranium release but substantially reduced the amount in solution by causing the uranium to plate out on the iron coupon and container walls and to precipitate as filterable particles.


1987 ◽  
Vol 112 ◽  
Author(s):  
P. L. Chambré ◽  
C. H. Kang ◽  
W. W.-L. Lee ◽  
T. H. Pigford

AbstractThe dissolution rate of waste solids in a geologic repository is a complex function of waste form geometry, chemical reaction rate, exterior flow field, and chemical environment. We present here an analysis to determine the steady-state mass transfer rate, over the entire range of flow conditions relevant to geologic disposal of nuclear waste. The equations for steady-state mass transfer with a chemical-reaction-rate boundary condition are solved by three different mathematical techniques which supplement each other. This theory is illustrated with laboratory leach data for borosilicate-glass and a spherical spent-fuel waste form under typical repository conditions. For borosilicate glass waste in the temperature range of 57°C to 250°C, dissolution rate in a repository is determined for a wide range of chemical reaction rates and for Peclet numbers from zero to well over 100, far beyond any Peclet values expected in a repository. Spent-fuel dissolution in a repository is also investigated, based on the limited leach data now available.


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