scholarly journals Solubility Limits. on Radionuclide Dissolution

1984 ◽  
Vol 44 ◽  
Author(s):  
Jerry F. Kerrisk
Keyword(s):  
Yucca Mountain ◽  
Waste Form ◽  
High Level Waste ◽  
Spent Fuel ◽  
Dissolution Rates ◽  
Diffusion Limited ◽  
Dissolution Model ◽  
High Level

AbstractThis paper examines the effects of solubility in limiting dissolution rates of a number of important radionuclides from spent fuel and high-level waste. Two simple dissolution models were used for calculations that would be characteristic of a Yucca Mountain repository. A saturation-limited dissolution model, in which the water flowing through the repository is assumed to be saturated with each waste element, is very conservative in that it overestimates dissolution rates. A diffusion-limited dissolution model, in which element-dissolution rates are limited by diffusion of waste elements into water flowing past the waste, is more realistic, but it is subject to some uncertainty at this time. Dissolution rates of some elements (Pu, Am, Sn, Th, Zr, Sm) are always limited by solubility. Dissolution rates of other elements (Cs, Tc, Np, Sr, C, I) are never solubility limited; their release would be limited by dissolution of the bulk waste form. Still other elements (U, Cm, Ni, Ra) show solubility-limited dissolution under some conditions.

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.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

scholarly journals Corrosion of Glass-Bonded Sodalite and its Components as a Function of ph and Temperature

2002 ◽  
Vol 713 ◽  
Author(s):  
Seung-Young Jeong ◽  
Lester R. Morss ◽  
William L. Ebert
Keyword(s):  
Dissolution Rate ◽  
Hot Isostatic Pressing ◽  
Waste Form ◽  
High Level Waste ◽  
Model Parameters ◽  
Dissolution Rates ◽  
Degradation Model ◽  
Modified Binder ◽  
Ph Range ◽  
High Level

ABSTRACTA glass-bonded sodalite ceramic waste form (CWF) has been developed to immobilize electrorefiner salt wastes from electrometallurgical treatment of spent sodium-bonded reactor fuel for disposal. A degradation model is being developed to support qualification of the CWF for disposal in the federal high-level waste disposal system. The parameter values in the waste form degradation model were previously determined from the dissolution rates measured in MCC-1 tests conducted at 40, 70, and 90°C. The results of several series of tests that were conducted to confirm the applicability of the dissolution rate model and model parameters are presented in this paper: (1) Series of MCC-1 tests were conducted in five dilute buffer solutions in the pH range of 4.8 – 9.8 at 20°C with hot isostatic pressing (HIP) sodalite, HIP glass, and HIP CWF. The results show that the model adequately predicts the dissolution rate of these materials at 20°C. (2) Tests at 20 and 70°C with CWF made by pressureless-consolidation (PC) indicate that the model parameters extracted from the results of tests with HIP CWF can be applied to PC CWF. (3) The dissolution rates of a glass made with a composition corresponding to 80 wt. % glass and 20 wt. % sodalite were measured at 70°C to evaluate the sensitivity of the rate to the composition of binder glass in the CWF. The dissolution rates of the modified binder glass were indistinguishable from the rates of the binder glass.

TRU Recycling Options for Environmentally Friendly and Proliferation-Resistant Nuclear Fuel Cycle of FBR

2010 ◽  
Author(s):  
Sidik Permana ◽  
Mitsutoshi Suzuki
Keyword(s):  
High Performance ◽  
Fuel Cycle ◽  
Operation Time ◽  
Nuclear Nonproliferation ◽  
High Level Waste ◽  
Minor Actinide ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
The Core ◽  
High Level

The embodied challenges for introducing closed fuel cycle are utilizing advanced fuel reprocessing and fabrication facilities as well as nuclear nonproliferation aspect. Optimization target of advanced reactor design should be maintained properly to obtain high performance of safety, fuel breeding and reducing some long-lived and high level radioactivity of spent fuel by closed fuel cycle options. In this paper, the contribution of loading trans-uranium to the core performance, fuel production, and reduction of minor actinide in high level waste (HLW) have been investigated during reactor operation of large fast breeder reactor (FBR). Excess reactivity can be reduced by loading some minor actinide in the core which affect to the increase of fuel breeding capability, however, some small reduction values of breeding capability are obtained when minor actinides are loaded in the blanket regions. As a total composition, MA compositions are reduced by increasing operation time. Relatively smaller reduction value was obtained at end of operation by blanket regions (9%) than core regions (15%). In addition, adopting closed cycle of MA obtains better intrinsic aspect of nuclear nonproliferation based on the increase of even mass plutonium in the isotopic plutonium composition.

scholarly journals The initial dissolution rates of simulated UK Magnox–ThORP blend nuclear waste glass as a function of pH, temperature and waste loading

2015 ◽  
Vol 79 (6) ◽  
pp. 1529-1542 ◽  
Author(s):  
N. Cassingham ◽  
C.L. Corkhill ◽  
D.J. Backhouse ◽  
R.J. Hand ◽  
J.V. Ryan ◽  
...  
Keyword(s):  
Dissolution Kinetics ◽  
Intrinsic Rate ◽  
Waste Glass ◽  
High Level Waste ◽  
Dissolution Mechanism ◽  
Dissolution Rates ◽  
Forward Rates ◽  
Glass Dissolution ◽  
High Level ◽  
Kinetics Of

AbstractThe first comprehensive assessment of the dissolution kinetics of simulant Magnox–ThORP blended UK high-level waste glass, obtained by performing a range of single-pass flow-through experiments, is reported here. Inherent forward rates of glass dissolution were determined over a temperature range of 23 to 70°C and an alkaline pH range of 8.0 to 12.0. Linear regression techniques were applied to the TST kinetic rate law to obtain fundamental parameters necessary to model the dissolution kinetics of UK high-level waste glass (the activation energy (Ea), pH power law coefficient (η) and the intrinsic rate constant (k0)), which is of importance to the post-closure safety case for the geological disposal of vitreous products. The activation energies based on B release ranged from 55 ± 3 to 83 ± 9 kJ mol–1, indicating that Magnox–THORP blend glass dissolution has a surface-controlled mechanism, similar to that of other high-level waste simulant glass compositions such as the French SON68 and LAW in the US. Forward dissolution rates, based on Si, B and Na release, suggested that the dissolution mechanism under dilute conditions, and pH and temperature ranges of this study, was not sensitive to composition as defined by HLW-incorporation rate.

scholarly journals Development of Integrated Mechanistically-Based Degradation-Mode Models for Performance Assessment of High-Level Waste Containers

1999 ◽  
Vol 556 ◽  
Author(s):  
J. C. Farmer ◽  
R. D. Mccright ◽  
J. C. Estill ◽  
S. R. Gordon
Keyword(s):  
Stainless Steel ◽  
Crevice Corrosion ◽  
Chloride Concentration ◽  
Yucca Mountain ◽  
304 Stainless Steel ◽  
High Level Waste ◽  
Mechanistic Models ◽  
Alloy 22 ◽  
High Level ◽  
Impurity Elements

AbstractAlloy 22 [UNS N06022] is now being considered for construction of high level waste containers to be emplaced at Yucca Mountain and elsewhere. In essence, this alloy is 20.0–22.5% Cr, 12.5–14.5% Mo, 2.0–6.0% Fe, 2.5–3.5% W, with the balance being Ni. Other impurity elements include P, Si, S, Mn, Co and V. Cobalt may be present at a maximum concentration of 2.5%. Detailed mechanistic models have been developed to account for the corrosion of Alloy 22 surfaces in crevices that will inevitably form. Such occluded areas experience substantial decreases in pH, with corresponding elevations in chloride concentration. Experimental work has been undertaken to validate the crevice corrosion model, including parallel studies with 304 stainless steel.

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