Evaluation of Spent Fuel as a Waste Form In a Salt Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
Walter J. GRAY ◽  
Gary L. McVay ◽  
John. O. Barner ◽  
John W. Shade ◽  
Roger W. Cote
Keyword(s):  
Deionized Water ◽  
Waste Form ◽  
Spent Fuel ◽  
Permian Basin ◽  
Salt Brine

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.

Mobilization/Retention of Radionuclides during Corrosion of High Burnup Spent Fuel and Backfill Materials in Salt Brines

2002 ◽  
Vol 757 ◽  
Author(s):  
Andreas Loida ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Near Field ◽  
Waste Form ◽  
Spent Fuel ◽  
The Matrix ◽  
Backfill Materials ◽  
Salt Brine ◽  
Solid Phases ◽  
Long Term Behavior ◽  
The Impact

ABSTRACTWith respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study the fuel alteration in contact with groundwater and near field materials. The aim of this work is to evaluate the impact of candidate backfill materials hydroxylapatite and magnetite on the overall corrosion behavior of this waste form in salt brine; both materials are used in corrosion tests together with spent fuel. The instant releases and the matrix dissolution rates appear to be similar in presence and in absence of any backfill material under study. However, Am,Np,Pu,U and Sr are retained at different ratios on the hydroxylapatite, on the magnetite and on the fuel sample, indicating possibly the formation of different radionuclide containing new solid phases.

Migration Rates of Brine Inclusions in Single Crystals of Nacl

1981 ◽  
Vol 6 ◽  
Author(s):  
I-Ming Chou
Keyword(s):  
Radioactive Waste ◽  
Temperature Gradients ◽  
Spent Fuel ◽  
Ion Diffusion ◽  
Cold Side ◽  
Migration Rates ◽  
Permanent Storage ◽  
Salt Brine ◽  
Salt Deposits ◽  
High Level

Rock-salt deposits have been considered as a possible medium for the permanent storage of high-level radioactive wastes and spent fuel. Brine inclusions present in natural salt can migrate toward the waste if the temperature and the temperature gradients in the vicinity of the radioactive waste are large enough. This migration is due to the dissolution of salt at the hot side of the salt-brine interface, ion diffusion through the brine droplet, and the precipitation of salt at the cold side of the salt brine interface.

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 ◽  
...  
Keyword(s):  
Production Process ◽  
Waste Form ◽  
Spent Fuel ◽  
Ceramic Waste

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

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.

scholarly journals The Role Of Chemical Reaction In Waste-Form Performance

1987 ◽  
Vol 112 ◽  
Author(s):  
P. L. Chambré ◽  
C. H. Kang ◽  
W. W.-L. Lee ◽  
T. H. Pigford
Keyword(s):  
Mass Transfer ◽  
Steady State ◽  
Dissolution Rate ◽  
Chemical Reaction ◽  
Borosilicate Glass ◽  
Reaction Rate ◽  
Waste Form ◽  
Spent Fuel ◽  
Wide Range ◽  
Chemical Reaction Rate

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.

Comparison of 200 Liter and 40 Milliliter Leach Tests

1982 ◽  
Vol 15 ◽  
Author(s):  
J. H. Westsik ◽  
C. O. Harvey ◽  
F. P. Roberts ◽  
W. A. Ross ◽  
R. E. Thornhill
Keyword(s):  
Large Scale ◽  
Deionized Water ◽  
Waste Form ◽  
Small Scale ◽  
Matrix Elements ◽  
Elemental Concentrations ◽  
The Past ◽  
Leach Test ◽  
The Matrix ◽  
Scale Test

ABSTRACTDuring the past year we have conducted a modified MCC-1 leach test on a 145 kg block of a cast cement waste form. The leach vessel was a 200 liter Teflon®-lined drum and contained 97.5 liters of deionized water. The results of this large-scale leach test were compared with the results of standard MCC-1 tests (40 ml) on smaller samples of the same waste form. The ratio of leachate volumes between the large and small scale tests was 2500 and the ratio of sample masses was 150,000. The cast cement samples for both tests contained plutonium-doped incinerator ash.The leachates from these tests were analyzed for both plutonium and the matrix elements. Evaluation of plutonium plateout in the large-scale test indicated that the majority of the plutonium leached from the samples deposits onto vessel walls and little (<3 × 10−12M) remains in solution. Comparison of elemental concentrations in the leachates indicates some differences up to 5X in the concentration in the large- and small-scale tests. The differences are attributed to differences in the solubilities of Ca, Si, and Fe at pH ˜11.5 and at pH ˜12.5. The higher pH observed for the large-scale test is a result of the larger quantities of sodium in the large block of cement.

scholarly journals The NNWSI Waste form Testing Program

1983 ◽  
Vol 26 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Release Rate ◽  
Chemical Elements ◽  
Performance Testing ◽  
Interactive Effects ◽  
Waste Form ◽  
Spent Fuel ◽  
Testing Program ◽  
Reactor Pool ◽  
Reaction Vessels ◽  
Topopah Spring Tuff

ABSTRACTA waste form testing program has been developed to ensure that the release rate of radionuclides from the engineered barrier system will meet NRC and EPA regulatory requirements. Waste form performance testing will be done under unsaturated, low water availability conditions which represent the expected repository conditions. Testing will also be done under conditions of total immersion of the waste form in repository-type water to cover the possibility that localized portions of the repository might contain standing water. Testing of reprocessed waste forms for CHLW and DHLW will use reaction vessels fabricated from Topopah Spring tuff. Chemical elements which are expected to show the highest release rates in the mildly oxidizing environment of the Topopah Spring tuff horizon at Yucca Mountain are Np and Tc. To determine the effect of residual canister material and of corrosion products from the canister/overpack, waste form testing will be done in the presence of these materials. The release rate of all radionuclides which are subject to NRC and EPA regulations will be measured, and the interactive effects of the released radionuclides and the rock reaction vessels will be determined. The testing program for spent fuel will determine the release rate from bare spent fuel pellets and from Zircaloy clad spent fuel where the cladding contains minor defects. A metal testing program for Zircaloy will establish the expected lifetime of the cladding material. Estimation of the state of cladding for fuel presently in reactor pool storage will provide baseline data for Zircaloy containment credit.

scholarly journals Using Center Hole Heat Transfer to Reduce Formation Times for Ceramic Waste Forms From Pyroprocessing

2006 ◽  
Author(s):  
Kenneth J. Bateman ◽  
Charles W. Solbrig
Keyword(s):  
Heat Transfer ◽  
Liquid Aluminum ◽  
Glass Frit ◽  
Waste Form ◽  
Electric Heater ◽  
Heating Process ◽  
Spent Fuel ◽  
Noticeable Effect ◽  
Flowing Liquid ◽  
Ceramic Waste

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.

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

2005 ◽  
Vol 346 (1) ◽  
pp. 24-31 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler ◽  
Horst Geckeis
Keyword(s):  
Spent Fuel ◽  
Salt Brine ◽  
High Burnup ◽  
Radionuclide Release
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