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.

Transient Diffusional Release from Waste Packages in a Repository in Basalt

1985 ◽  
Vol 50 ◽  
Author(s):  
M. S. Bensky ◽  
D. L. Oliver
Keyword(s):  
Release Rate ◽  
Host Rock ◽  
Near Field ◽  
Temporal Variations ◽  
Waste Form ◽  
Spent Fuel ◽  
Constant Concentration ◽  
Release Rates ◽  
Total Inventory ◽  
System Characteristics

AbstractAnalyses of diffusional release of several typical radionuclides in spent fuel from waste packages emplaced in a repository in basalt were conducted to assess the effects of system characteristics and boundary conditions on computed release rates. Radionuclide releases, including spatial and temporal variations that may be present, represent the source term for transport in the geohydrologic setting and are therefore critical to the assessment of repository acceptability.Two mathematical approaches were utilized to determine radionuclide release rate versus time characteristics; (1) an analytical solution for one-deimensional diffusion based upon a Dirichlet (constant-concentration) boundary at the waste form surface; and (2) a finite-element numerical solution based upon a Neumann (zero-flux boundary at the waste form surface. The latter method is suitable for radionuclides such as 129I, whose total inventory in spent fuel could be quickly depleted from the waste form and dissolved in the pore spaces of the packing material surrounding the waste form and which, therefore, cannot be adequately represented by a constant concentration at the waste form (i.e., container) surface.The analysis revealed several system characteristics that are not intuitively obvious. For example, strong sorption in the near-field host rock behaves like a strong mass sink and can yield calculated transient release rates exceeding allowable limits. Similarly, a short half-life effectively removes the radionuclide from the host rock, which induces a steep concentration gradient at the host rock/packing interface and thereby increases the diffusional release rate at that boundary.Typical results for 79Se and 129I are presented to illustrate these effects. The effects of perturbations to key assumptions are shown to indicate the importance of (1) formulating models that accurately represent the physical system and (2) interpreting analytical results carefully to ensure understanding of the capability of the system.

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

Engineered Components for Spent Fuel Radioactive Waste Isolation Systems—are They Technically Justified?

1981 ◽  
Vol 11 ◽  
Author(s):  
H. C. Burkholder
Keyword(s):  
United States ◽  
Radioactive Waste ◽  
Release Rate ◽  
The United States ◽  
Radioactive Waste Disposal ◽  
Radionuclide Transport ◽  
Spent Fuel ◽  
Isolation System ◽  
Waste Forms ◽  
Isolation Systems

In response to draft radioactive waste disposal standards, R&D programs have been initiated in the United States which are aimed at developing and ultimately using radionuclide transport-delaying (e.g., long-lived waste containers) and radionuclide transport-controlling (e.g., very low release rate waste forms) engineered components as part of the isolation system. Before these programs proceed significantly, it seems prudent to evaluate the technical justification for development and use of sophisticated engineered components in radioactive waste isolation.

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.

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.

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.

Leaching of Spent Fuel under Anaerobic and Reducing Conditions

2002 ◽  
Vol 757 ◽  
Author(s):  
Yngve Albinsson ◽  
Arvid Ödegaard-Jensen ◽  
Virginia M. Oversby ◽  
Lars O. Werme
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Fluid Phase ◽  
Inert Atmosphere ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Reducing Conditions ◽  
Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

scholarly journals A Proficiency Testing Program for the Agricultural Laboratory Analysis Industry

HortScience ◽  
1997 ◽  
Vol 32 (3) ◽  
pp. 456G-457
Author(s):  
Robert O. Miller ◽  
Steven E. Newman ◽  
Janice Kotuby-Amacher
Keyword(s):  
Proficiency Testing ◽  
Performance Testing ◽  
Internal Quality ◽  
Nitrogen And Phosphorus ◽  
Testing Program ◽  
Plant Nitrogen ◽  
Plant Samples ◽  
Accuracy And Precision ◽  
Internal Quality Assurance ◽  
Canadian Provinces

The accuracy of soil and plant analytical results are occasionally called into question by laboratory clientele. Although laboratories generally conduct internal quality assurance procedures, there are few external performance testing programs for the industry. In 1994, a proficiency testing program was initiated for soil and plant samples for agricultural laboratories in the western United States to provide an external quality control for the lab industry. The program involves the quarterly exchange of soil and plant samples on which soil salinity, soil fertility, and plant nutrition analyses are conducted. One hundred laboratories are annually enrolled in the program from 24 states and Canadian provinces. Results of 3 years of the program indicate soil nitrate, soil pH, extractable potassium, soil and organic matter are reproducible within 10% between laboratories. Soil-extractable phosphorus (by five methods), soil-extractable boron, and soluble chloride were only reproducible within 15% to 20% between laboratories. Plant nitrogen and phosphorus results were consistent across samples, laboratories, and methods. Variability in plant nitrate increased with decreasing tissue concentrations. Overall accuracy and precision of reported results, based on the use of NIST certified reference botanical samples, were excellent for N, P, K, Ca, and Cu. Generally, for any given analysis, the results of ≈10% of the laboratories exceed two standard deviations from the mean. Overall, significant improvement was noted in the laboratory industry proficiency through the course of the program.

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.

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