Effect of Heat Treatment on Corrosion Behavior of Alloy 690 and Alloy 693 in Simulated Nuclear High-Level Waste Medium

CORROSION ◽  
2012 ◽  
Vol 68 (4) ◽  
pp. 046001-1-046001-13 ◽  
Author(s):  
P.K. Samantaroy ◽  
S. Girija ◽  
U. Kamachi Mudali
Keyword(s):  
Corrosion Behavior ◽  
Current Density ◽  
Spent Nuclear Fuel ◽  
Potassium Thiocyanate ◽  
Solution Temperature ◽  
Annealed Specimen ◽  
High Level Waste ◽  
Alloy 690 ◽  
Passivation Current Density ◽  
High Level

Nickel-based alloys are being considered as candidate materials for the storage of high-level waste generated from the reprocessing of spent nuclear fuel. In the present investigation, Alloy 690 (UNS N06690) and Alloy 693 (UNS N06693) were assessed using the potentiodynamic anodic polarization technique for the corrosion behavior in the solution-annealed and sensitized conditions, in 3 M nitric acid (HNO3) and simulated high-level waste (HLW) at 25°C and 50°C. From the results of the investigation, it was found that the polarization curves for the solution-annealed specimens were characterized by a long passive range and low passivation current density, at 25°C. Increasing solution temperature to 50°C led to a corresponding increase in corrosion potential as well as passivation current density. The solution-annealed specimen showed improved corrosion resistance compared to the sensitized specimen in both media, for both alloys. Double-loop electrochemical potentiokinetic reactivation (DL-EPR) test was carried out for both solution-annealed and sensitized Alloy 690 and Alloy 693 in 0.5 M sulfuric acid (H2SO4) containing 0.0001 M potassium thiocyanate (KSCN), to measure the degree of sensitization (DOS). Microstructural examination was carried out by optical microscopy and scanning electron microscopy (SEM) after electrolytic etching. The results of the present investigation are discussed in the paper.

Corrosion behavior of Alloy 690 and Alloy 693 in simulated nuclear high level waste medium

2011 ◽  
Vol 418 (1-3) ◽  
pp. 27-37 ◽  
Author(s):  
Pradeep Kumar Samantaroy ◽  
Girija Suresh ◽  
Ranita Paul ◽  
U. Kamachi Mudali ◽  
Baldev Raj
Keyword(s):  
Corrosion Behavior ◽  
High Level Waste ◽  
Alloy 690 ◽  
High Level

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 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.

Radionuclide Uptake and Transport on Microbes in Potential Repository Drifts at Yucca Mountain, Nevada

2002 ◽  
Vol 713 ◽  
Author(s):  
Darren M. Jolley
Keyword(s):  
Source Term ◽  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Multiple Model ◽  
Data Set ◽  
Model Calculations ◽  
Waste Package ◽  
Uranium Uptake ◽  
High Level ◽  
Dry Cell

ABSTRACTRadionuclide adsorption onto microbes, microbial retention in the engineered barrier system (EBS), and their potential release from the EBS as microbial colloids have been investigated. The microbial source term for these calculations was derived using MING V 1.0 software code [1]. Multiple model calculations from MING representing variations in possible microbial communities in the EBS were abstracted into two equations representing one meter segments of potential repository drift containing either commercial spent nuclear fuel (CSNF) or defense high level waste (HLW) packages. These two equations (Equations 1 and 2) represent the average cumulative microbial biomass generated in the EBS at any given time. A distribution for uranium uptake onto microbes (162.88 ± 133.05 mg U/gm dry cell) was applied to the microbial source term. The distribution was derived from the data set in Suzuki and Banfield [2] representing 45 different species of bacteria and fungus, covering uranium uptake at optimum pH values of 1 to 7. The mass of uranium sorbed onto the biomass was either sequestered in the EBS or transported as a microbial colloid based on a regression of data from Jewett et al. [3] representing microbial sorption onto air-water interfaces in unsaturated column experiments. Over one million years, it is estimated that EBS microbes may adsorb from 77 to 2302 kg of uranium [2302 kg U > 100% of the uranium available in a one meter segment of a CSNF waste package] per meter of waste package depending on the saturation of the invert and type of waste package. Over the same time, microbial colloids may transport from 8 to 1250 kg of adsorbed uranium per meter of waste package from the EBS.

scholarly journals Development of a ceramic waste form for high-level waste disposal

1999 ◽  
Vol 556 ◽  
Author(s):  
D. W. Esh ◽  
K. M. Goff ◽  
K. T. Hirsche ◽  
T. J. Battisti ◽  
M. F. Simpson ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
High Level Waste ◽  
Durability Test ◽  
Ceramic Waste ◽  
High Level ◽  
The Impact

AbstractA ceramic waste form is being developed by Argonne National Laboratory* (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel [1]. The halide, alkaline earth, alkali, transuranic, and rare earth fission products are stabilized in zeolite which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The mineral sodalite is formed in the HIP from the zeolite precursor. The process, from starting materials to final product, is relatively simple. An overview of the processing operations is given. The metrics that have been developed to measure the success or completion of processing operations are developed and discussed. The impact of variability in processing metrics on the durability of the final product is presented. The process is demonstrated to be robust for the type and range of operation metrics considered and the performance metric (PCT durability test) against which the operation metrics are evaluated.

Characteristics of Cementitious Paste for use in Deep Borehole Disposal of Spent Fuel and High Level Wasteforms

2015 ◽  
Vol 1744 ◽  
pp. 205-210 ◽  
Author(s):  
Nick C Collier ◽  
Karl P Travis ◽  
Fergus G F Gibb ◽  
Neil B Milestone
Keyword(s):  
Spent Nuclear Fuel ◽  
High Level Waste ◽  
Hardened Cement ◽  
Oil Well ◽  
Deep Borehole ◽  
Sodium Gluconate ◽  
Spent Fuel ◽  
High Temperature And Pressure ◽  
Temperature And Pressure ◽  
High Level

ABSTRACTDeep borehole disposal (or DBD) is now seen as a viable alternative to the (comparatively shallow) geologically repository concept for disposal of high level waste and spent nuclear fuel. Based on existing oil and geothermal well technologies, we report details of investigations into cementitious grouts as sealing/support matrices (SSMs) for waste disposal scenarios in the DBD process where temperatures at the waste package surface do not exceed ∼190ºC. Grouts based on Class G oil well cements, partially replaced with silica flour, are being developed, and the use of retarding admixtures is being investigated experimentally. Sodium gluconate appears to provide sufficient retardation and setting characteristics to be considered for this application and also provides an increase in grout fluidity. The quantity of sodium gluconate required in the grout to ensure fluidity for 4 hours at 90, 120 and 140°C is 0.05, 0.25 and 0.25 % by weight of cement respectively. A phosphonate admixture only appears to provide desirable retardation properties at 90°C. The presence of either retarder does not affect the composition of the hardened cement paste over 14 days curing and the phases formed are durable under conditions of high temperature and pressure.

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