NMR Study of Interlayer and Non-interlayer Porewater in Water-saturated Bentonite and Montmorillonite

2013 ◽  
Vol 1518 ◽  
pp. 167-172
Author(s):  
Torbjörn Carlsson ◽  
Arto Muurinen ◽  
Andrew Root
Keyword(s):  
Spent Nuclear Fuel ◽  
Relative Volume ◽  
High Level Waste ◽  
Proton Relaxation ◽  
Dry Density ◽  
Nmr Study ◽  
Important Barrier ◽  
Relaxation Curves ◽  
High Level ◽  
Water Saturated

ABSTRACTBentonite is planned to be used in many countries as an important barrier in high-level waste repositories. Assessment of the barrier with regard to, inter alia, its ability to hinder transport of dissolved radionuclides leaking from a damaged canister containing spent nuclear fuel, requires quantitative data about the pore structure inside bentonite. The present NMR study was made in order to determine the number of distinguishable porewater phases in compacted water-saturated samples of MX-80 bentonite and Na-montmorillonite. The samples were compacted to dry densities in the interval 0.7-1.6 g/cm3 and subsequently saturated with Milli-Q water or 0.1 M NaCl solution in equilibrium cells. The NMR measurements were performed with a high-field 270 MHz NMR spectrometer using a short inter-pulse CPMG method to study proton T1ρ relaxation. The measured relaxation curves were found to consist of one faster and one slower proton relaxation. Subsequent analysis of the data indicated that the faster relaxation was associated with interlayer (IL) water between montmorillonite unit layers, while the slower one was associated with non-interlayer (non-IL) water located outside the interlayer spaces. The results indicate for compacted samples with a dry density of ≥ 1.0 g/cm3, that Na montmorillonite contains a larger relative volume of non-IL water than the corresponding MX-80 bentonite. This in turn, suggests that the stacking number in Na-montmorillonite is smaller than in MX-80 bentonite. Changing the porewater chemistry seemed to have some effect on the non-IL water content in the Na montmorillonite but not in the MX-80 bentonite.

scholarly journals Comparison of microstructural features of three compacted and water-saturated swelling clays: MX-80 bentonite and Na- and Ca-purified bentonite

Clay Minerals ◽  
2019 ◽  
Vol 54 (1) ◽  
pp. 75-81 ◽  
Author(s):  
Michał Matusewicz ◽  
Markus Olin
Keyword(s):  
Spent Nuclear Fuel ◽  
Bulk Material ◽  
Fine Fraction ◽  
High Level Waste ◽  
Cation Composition ◽  
Swelling Clays ◽  
Transmission Electron ◽  
And Performance ◽  
High Level ◽  
Water Saturated

AbstractThe planned final disposal repositories of spent nuclear fuel in several countries, including Finland, pose significant scientific challenges due to their extremely long lifespan. One of the key materials proposed for use in Posiva Oy's repository in Finland is MX-80 bentonite in a compacted, water-saturated state. Border cases of calcium and sodium forms of purified bentonite were included in this study. The MX-80 in the repository is expected to undergo cation exchange due to the composition of the groundwater. The clays were studied at different dry densities between 0.7 and 1.6 g cm–3. The microstructure of the water-saturated, compacted clays was investigated using small-angle X-ray scattering, nuclear magnetic resonance and transmission electron microscopy. Additionally, atomic force microscopy was used to characterize the shape and size of the fine-fraction clay platelets. As expected, the average shape of the fine fractions was smaller than the bulk material, but a more elongated shape was present in the purified material. Mainly due to sample density, the pore structure was noticeably different for the Na form of purified bentonite at 0.7 g cm–3 density, but at higher degrees of compaction, no significant differences were noted between the samples. The laboratory results obtained in this study could be useful for safety and performance analysis of this high-level waste repository where sodium bentonite is used and is expected to change its charge-compensating cation composition during the repository's lifetime. Microstructural data may be used in modelling of diffusion and sorption by surface complexation modelling, for example, or as a basis for mechanical and water transport models.

A resistivity-based approach to determining the rates of groundwater seepage into buffer materials

2019 ◽  
Vol 482 (1) ◽  
pp. 205-212 ◽  
Author(s):  
T. Ishii ◽  
M. Kawakubo ◽  
H. Asano ◽  
I. Kobayashi ◽  
P. Sellin ◽  
...  
Keyword(s):  
Water Solution ◽  
High Level Waste ◽  
Dry Density ◽  
Chemical Alteration ◽  
Constant Rate ◽  
Buffer Material ◽  
The Status ◽  
Engineered Barriers ◽  
High Level ◽  
Saturation Front

AbstractBentonite-based buffer materials play an important safety role in engineered barriers planned for use in geological disposal repositories for radioactive high-level waste (HLW) in Japan. The effectiveness of buffer materials is dependent on the status of groundwater saturation during resaturation of the repository. Accordingly, it is important to determine the behaviour of buffer materials during saturation and predict post-saturation conditions such as the distribution of residual dry density and chemical alteration.In this study, the rate of groundwater uptake into a buffer material was determined to clarify the behaviour of the material during the saturation process. As mechanical changes and chemical alteration of buffer materials are generated by groundwater permeation, knowledge of the water uptake rate is necessary for the prediction of post-permeation conditions. In the experiment reported here, one-dimensional permeation by distilled water and a NaCl water solution at a constant rate was monitored over a period of more than seven years. The results indicated that the seepage and saturation front moved in proportion to the square root of the seepage time. The coefficient of the relationships between the seepage and the saturation fronts with time of the reference bentonite used in Japan was determined.

scholarly journals Reactive transport calculations to evaluate sulphide fluxes in the near-field of a SF/HLW repository

2019 ◽  
Vol 98 ◽  
pp. 10005
Author(s):  
Marek Pękala ◽  
Paul Wersin ◽  
Veerle Cloet ◽  
Nikitas Diomidis
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Waste Containment ◽  
Geological Repository ◽  
High Level ◽  
Reducing Bacteria ◽  
Hlw Repository

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Cu coating of the steel disposal canister is considered as potential a measure to ensure complete waste containment of spent nuclear fuel (SF) and vitrified high-level waste (HLW) or a period of 100,000 years. Sulphide is a potential corroding agent to Cu under reducing redox conditions. Background dissolved sulphide concentrations in pristine OPA are low, likely controlled by equilibrium with pyrite. At such concentrations, sulphide-assisted corrosion of Cu would be negligible. However, the possibility exists that sulphate reducing bacteria (SRB) might thrive at discrete locations of the repository’s near-field. The activity of SRB might then lead to significantly higher dissolved sulphide concentrations. The objective of this work is to employ reactive transport calculations to evaluate sulphide fluxes in the near-field of the SF/HLW repository in the OPA. Cu canister corrosion due to sulphide fluxes is also simplistically evaluated.

Adsorption Materials Development for the Separation of Actinides and Specific Fission Products from High Level Waste

2014 ◽  
Vol 94 ◽  
pp. 103-110 ◽  
Author(s):  
Yue Zhou Wei ◽  
Shun Yan Ning ◽  
Qi Long Wang ◽  
Zi Chen ◽  
Yan Wu ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Selective Adsorption ◽  
Packed Column ◽  
Liquid Waste ◽  
Separation Method ◽  
High Level Waste ◽  
Small Scale ◽  
Fast Diffusion ◽  
High Level Liquid Waste ◽  
High Level

The long-term radiotoxicity of high level liquid waste (HLLW) generated in spent nuclear fuel reprocessing is governed by the content of several long-lived minor actinides (MA) and some specific fission product nuclides. To efficiently separate MA (Am, Cm) and some FPs such as Cs and Sr from the HLLW, we have been studying an advanced aqueous partitioning process, which uses selective adsorption as separation method. In this work, we prepared different types of porous silica-based organic/inorganic adsorbents with fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. So they are advantageously applicable to efficient separation of the MA and specific FP elements from HLLW. Adsorption and separation behaviors of the MA and some FP elements such as Cs and Sr were studied. Small scale separation tests using simulated and genuine nuclear waste solutions were carried out and the obtained results indicate that the proposed separation method based on selective adsorption is essentially feasible.

scholarly journals Sorption-Capacity Limited Retardation of Radionuclides Transport in Water-Saturated Packing Materials

1984 ◽  
Vol 44 ◽  
Author(s):  
C. Pescatore ◽  
T. Sullivan
Keyword(s):  
Sorption Capacity ◽  
Retardation Factor ◽  
High Level Waste ◽  
Radionuclide Transport ◽  
Breakthrough Time ◽  
Packing Materials ◽  
Sorption Data ◽  
Waste Package ◽  
High Level ◽  
Water Saturated

AbstractRadionuclides breakthrough times as calculated through constant retardation factors obtained in dilute solutions are non-conservative. The constant retardation approach regards the solid as having infinite sorption capacity throughout the solid. However, as the solid becomes locally saturated, such as in the proximity of the waste form-packing materials interface, it will exhibit no retardation properties, and transport will take place as if the radionuclides were locally non-reactive. The magnitude of the effect of finite sorption capacity of the packing materials on radionuclide transport is discussed with reference to high-level waste package performance. An example based on literature sorption data indicates that the breakthrough time may be overpredicted by orders of magnitude using a constant retardation factor as compared to using the entire sorption isotherm to obtain a concentrationdependent retardation factor.

scholarly journals Characterization of Composite Ceramic High Level Waste Forms

1997 ◽  
Vol 481 ◽  
Author(s):  
S. M. Frank ◽  
K. J. Bateman ◽  
T. DiSanto ◽  
S. G. Johnson ◽  
T. L. Moschetti ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Composite Ceramic ◽  
Waste Form ◽  
High Level Waste ◽  
Ceramic Waste ◽  
Chloride Form ◽  
Physical Measurements ◽  
High Level

ABSTRACTArgonne National Laboratory has developed a composite ceramic waste form for the disposition of high level radioactive waste produced during electrometallurgical conditioning of spent nuclear fuel. The electrorefiner LiCl/KCl eutectic salt, containing fission products and transuranics in the chloride form, is contacted with a zeolite material which removes the fission products from the salt. After salt contact, the zeolite is mixed with a glass binder. The zeolite/glass mixture is then hot isostatic pressed (HIPed) to produce the composite ceramic waste form. The ceramic waste form provides a durable medium that is well suited to incorporate fission products and transuranics in the chloride form. Presented are preliminary results of the process qualification and characterization studies, which include chemical and physical measurements and product durability testing, of the ceramic waste form.

Corrosion Resistance of Iron-Based Amorphous Metal Coatings

2006 ◽  
Author(s):  
J. C. Farmer ◽  
J. J. Haslam ◽  
S. D. Day ◽  
T. Lian ◽  
R. Rebak ◽  
...  
Keyword(s):  
Corrosion Resistance ◽  
Thermal Spray ◽  
Spent Nuclear Fuel ◽  
Amorphous Metal ◽  
High Level Waste ◽  
Metal Coatings ◽  
Spray Coatings ◽  
Type 316L ◽  
High Level ◽  
Type 316L Stainless Steel

New amorphous-metal thermal-spray coatings have been developed recently that may provide a viable coating option for spent nuclear fuel & high-level waste repositories [Pang et al. 2002; Shinimiya et al. 2005; Ponnambalam et al. 2004; Branagan et al. 2000–2004]. Some Fe-based amorphous-metal formulations have been found to have corrosion resistance comparable to that of high-performance alloys such as Ni-based Alloy C-22 [Farmer et al. 2004–2006]. These materials rely on Cr, Mo and W for enhanced corrosion resistance, while B is added to promote glass formation and Y is added to lower the critical cooling rate (CCR). Materials discussed in this paper include yttrium-containing SAM1651 with CCR ∼ 80 K/s and yttrium-free Formula 2C with CCR ∼ 600 K/s. While nickel-based Alloy C-22 and Type 316L stainless steel lose their resistance to corrosion during thermal spraying, Fe-based SAM1651 and Formula 2C amorphous-metal coatings can be applied with thermal spray processes without any significant loss of corrosion resistance. In the future, such corrosion-resistant thermal-spray coatings may enable the development of less expensive containers for spent nuclear fuel (SNF) and high-level waste (HLW), including enhanced multipurpose containers (MPCs), protected closure welds, and shields to protect containers from drips and falling rocks. These materials are extremely hard and provide enhanced resistance to abrasion and gouges from backfill operations. For example, Type 316L stainless steel has a hardness of approximately 150 VHN, Alloy C-22 has a hardness of approximately 250 VHN, while the Fe-based amorphous metals typically have hardness values of 1100–1300 VHN. Both Formula 2C and SAM1651 have high boron content which allow them to absorb neutrons, and therefore be used for enhanced criticality control. Cost savings can also be realized through the substitution of Fe-based alloy for Ni-based materials. Applications are also envisioned in oil & gas industry.

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