Carbon Steel Containers Research Articles

Interaction of 99Tc onto bentonite and influence of Fe(II) and (Ca, Mg) ion exchange in its interlayers

99Tc is a redox active radionuclide, which is present as contaminant at a number of sites where nuclear fuel cycle operations have been carried out. The aim of our research was study of bentonite interaction with Fe and Fe2+ cation and these influences on the \( ^{ 9 9} {\text{TcO}}_{4}^{ - } \) migration behavior. Radioanalytical and analytical methods were used for the concentration and chemical forms of 99Tc determination. It was found that Fe2+–Ca2+, Mg2+ ion exchange should be considered as a process of reaction among the corrosion products of carbon steel container and bentonite in the environment of radioactive waste repository.

Origin of the hydrogen involved in iron corrosion under irradiation

In the perspective of long term geological storage, high level nuclear wastes will be overpacked in low carbon steel containers. In that context, we have studied the influence of oxygen dissolved in water on iron corrosion. Therefore, leaching experiments were performed in deaerated D 2O and in aerated H 2O and a kinetic study of iron corrosion under proton irradiation was lead in aqueous media with two different dissolved oxygen concentrations. The leaching experiments underline the major role of dissolved oxygen in oxydoreduction reactions which take place as far as iron is in contact with water. But the kinetic study of iron corrosion under irradiation put in evidence the balance between the oxydoreduction reactions and the corrosion rate induced by radical species generated by water radiolysis. In addition, to check if, in the atomic % concentration range, hydrogen diffuse from the air/Fe interface through the foil, an irradiation experiment was performed in argon. It proved that no hydrogen permeation occurs at a concentration level of the atomic percent.

Toughness of Welded Low Carbon Steels Containing Nitrogen (A Resolved Problem on Recycled Steels)

Toughness of Welded Low Carbon Steels Containing Nitrogen (A Resolved Problem on Recycled Steels)

TEM Characterization of the Recrystallization Behaviour of Warm Rolled Low Carbon Steels Containing Chromium during the Early Stages of Annealing

The early stages of recovery and recrystallisation were studied using transmission electron microscopy in four warm rolled low carbon steels. Three of these contained additions of chromium and of the latter, one was phosphorus and a second boron modified. Addition of the alloying elements led to the formation of both shear and microbands within some of the grains. The progress of recovery in these areas differed from that applicable to the grains containing only microbands. Two types of carbides were present after warm rolling in the steels containing the alloying additions: (i) coarse carbides; and (ii) fine strain-induced particles. The coarse carbides underwent spheroidization and coarsening during annealing. They, on the one hand, accelerated recrystallisation as the particles stimulated nucleation; on the other hand, they retarded it by pinning both the high and low angle grain boundaries. These carbides also influenced the morphology of the recrystallised grains by restraining their growth.

Fe(II)–Na ion exchange at interlayers of smectite: adsorption–desorption experiments and a natural analogue

Experiments of Fe 2+ adsorption–desorption to Na-smectite were carried out to understand the effects of the corrosion products of a carbon-steel container on the alteration of bentonite as a buffer material in the environment of an underground repository for radioactive waste. Refined smectite reacted with Fe 2+ in NaCl solutions under conditions that proportions of the liquid/solid (L/S=solution/smectite) ranged from 50 to 500 ml/g, and the ionic strengths ( I) were 0.001 and 0.05. Ion selectivity coefficients ( K GT) for the reaction of 2XNa+Fe 2+=X 2Fe+2Na + were obtained: 0.06±0.00 when I=0.001, L/S=200 ml/g, and 1.85±0.29 when I=0.05, L/S=50 ml/g. A natural analogue of the alteration between a carbon steel container and bentonite, where an iron-support has been kept in contact with Na-type bentonite for about 2000 days at a gallery in a bentonite mine, was also studied. It was found that the color of the bentonite had changed from pale gray to green near the iron support. The redox potential ( E h) could be measured directly for the ‘green bentonite’ and was found to be from −8 to +46 mV vs. SHE. Powder XRD analyses for the ‘green bentonite’ showed that a variation of d 001-reflections with changing ambient relative humidity was in fairly good agreement with that of a divalent cation adsorbed smectite. Exchange between Fe 2+and Na + at interlayers of smectite could therefore occur in a natural reductive environment.

Open Site Tests on Corrosion of Carbon Steel:Containers for Radioactive Waste Forms

AbstractTesting of waste containers under open field conditions is a component part of the research program that is being carried out at SIA “Radon”for more than 20 years to understand the long-term behaviour of radioactive waste forms and waste packages. This paper presents the preliminary results of these ongoing studies. We used a typical NPP operational waste, containing 137Cs, 134Cs, and 60Co as the dominant radioactive constituents. Bituminized and vitrified waste samples with 30–50 wt.% waste loading were prepared. Combined effects of climatic factors on corrosion behaviour of carbon steel containers were estimated using gravimetric and chemical analyses. The observations suggest that uniform corrosion of containers prevails under open field conditions. The upper limits for the lifetime of containers were derived from calculations based on the model of atmospheric steel corrosion. Estimated lifetime values range from 300 to 600 years for carbon steel containers with the wall thickness of 2 mm containing vitrified waste, and from 450 to 500 years for containers with the wall thickness of 2.5 mm that were used for bituminized waste. However, following the most conservative method, pitting corrosion may cause container integrity failure after 60 to 90 years of exposure.

Corrosion of Carbon Steel Containers During Open Site Tests of Vitrified Radioactive Waste

ABSTRACTIn the middle of 80-s the Moscow SIA “Radon” initiates a program on long term tests of vitrified intermediate level radioactive waste from nuclear power plants with different types'of reactors and with different composition of waste. Glass blocks were prepared and disposed for tests in experimental shallow ground repository and on open testing area. Borosilicate glass as matrix material is investigated in order to obtain data on real behaviour of glass [1]. The waste glass was poured in carbon steel rectangular containers therefore containers also were subjected to aqueous and atmospheric corrosion. Exposure conditions at testing site are varying depending on season. They can be described only as average parameters with some uncertainties.

Lead content in grapefruit juice and its uptake upon storage in open containers

AbstractThe lead contents of 52 samples of grapefruit juice, 14 packaged in glass, eight in waxed paperboard, and 30 in tin‐coated carbon steel containers, were determined. Only those juices packaged in metal cans contained measurable amounts of lead, (< 1–27 μg litre‐1) averaging 7‐7 μg litre‐1. Leaching studies were carried out on juices stored, under refrigeration, in opened containers for up to 30 days. The lead content increased considerably with time in juices stored in tin‐coated cans. For example, after 6 days of storage the lead contents of these juices ranged from 3 to 90 μg litre‐1, and averaged 30 μg litre‐1. The lead content did not increase in juices that were stored in glass, paperboard or polyethylene containers. The source of lead in the cans was traced to lead impurities in the tin coating. Higher lead impurities in the tin coating produced higher lead contents in the juice. The implications of these findings are discussed.

Corrosion of High-Level Waste Packaging Materials in Disposal Relevant Brines

Previous corrosion studies identified the materials Hastelloy C4, Ti 99.8-Pd, and carbon steels as promising for the manufacture of long-lived high-level waste containers that could act as an engineered barrier in a rock-salt repository. Here, the efficiency of the corrosion-resistant concept using surface-welded Hastelloy C4 as corrosion protection of carbon steel containers is compared with the corrosion-allowance concept using unalloyed or low-alloyed steels. The materials are examined in three disposal relevant brines (two rich in MgCl2, one rich in NaCl) at 150°C. The results indicate that welded Hastelloy C4 is highly resistant to corrosion in the NaCl-rich brine. In the presence of sulfides or MgCl2-rich brines, however, severe pitting corrosion occurs. The three steels investigated are resistant to pitting corrosion in all brines, and their general corrosion rates imply corrosion allowances acceptable for thick-walled containers. In view of these results, carbon steels continue to be considered promising materials for long-lived containers. Further investigations on carbon steels and Ti 99.8-Pd as alternatives to Hastelloy C4 are in progress.

Experiments and Modelling Studies Concerning Localised Corrosion of Carbon Steel and Stainless Steel Containers for Intermediate- and Low-Level Radioactive Waste

ABSTRACTCurrent plans for disposal of low- and intermediate-level radioactive wastes (LLW and ILW) in the UK include enclosing much of the waste in carbon steel or stainless steel containers. Modelling and experimental studies that aim to characterise the period and extent of the localised corrosion in these containers are described. The period, during which localised corrosion can be initiated and sustained in the post-closure phase of the repository is estimated. The likely modes of localised corrosion are identified, based on further consideration of the environmental conditions to which the metals are exposed. Detailed research in progress includes investigation of the rate of pitting corrosion for carbon steel, and the possible occurrence of crevice corrosion of stainless steel.

Corrosion of Carbon-Steel Containers for Heat-Generating Nuclear Waste in Brine Environments Relevant for A Rock-Salt Repository

ABSTRACTIn previous corrosion studies, carbon steels, especially the fine-grained steel TStE355, were identified as promising materials for heat-generating nuclear waste containers acting as a barrier in a rock-salt repository. In the present study detailed investigations have been performed on fine-grained steel to determine the influence of important parameters on its corrosion behaviour in disposal-relevant salt brines. These parameters are: brine composition (Mg Cl2-rich and NaCI-rich brines), temperature (90°C, 170°C), and salt impurities, such as H2S concentrations of 25 mg/I-200 mg/I salt brine.Under the conditions of the tests used here, carbon steel was subjected to general corrosion. Pitting and crevice corrosion or stress-corrosion cracking were not observed. The increase in temperature from 90°C to 170°C strongly enhanced the corrosion rate of the steel. In the MgCl2-rich brines, considerably higher rates (37-70 μm/a at 90°C, 200-300 °m/a at 170°C) were observed than in the NaCI-rich brine (5 μm/a at 90°C, 46 μm/a at 170°C). H2S concentrations in the MgCl2-rich Qbrine of up to 200 mg/l did not influence significantly the corrosion rate of the steel. The corrosion rates determined imply corrosion allowances that are technically acceptable for thick-walled containers. In view of these results, fine-grained steel continues to be considered as a promising material for long-lived HLW containers.

In-Situ Corrosion Studies on Cast Steel High-Level Waste Containers Plated with Titanium/Nickel-Alloys

ABSTRACTPrevious corrosion studies have shown that the passive alloys Ti 99.8-Pd and Hastelloy C4 are the most promising corrosion protection materials for long-term resistant carbon-steel containers for the disposal of heat-generating wastes in rock salt formations. Consequently, these materials are subject to more detailed investigations. The influence of selected container manufacturing characteristics (sealing technique, application mode of the corrosion protecting layer on the steel) on the corrosion behaviour of Ti 99.8-Pd and Hastelloy C4was studied using in-situ experiments. The corrosion resistance of the materials was examined both in NaCI brine and MgCl2-rich brine, which might enter into an HLW borehole under certain disposal conditions. Four electron-beam welded cast-steel tubes plated with corrosion protection layers made of Ti 99.8-Pd and Hastelloy C4, respectively, were stored for 18 months in 2-m deep heated boreholes in the Asse salt mine. The annular gap between the tubes and the borehole wall was filled with saturated NaCl-brine or Qbrine (MgCl2-rich). The vertical temperature profile in the boreholes ranged from 90°C to 200°CIn the cast-steel tube with a corrosion protection of Ti 99.8-Pd, no indication of corrosion was observed on the EB-welds or on the explosion plated material in both brines. The cast-steel tube with a corrosion protection of Hastelloy C4 also was found to be resistant to any kind of corrosion in the NaCl-brine. In the MgCl2-rich brine, however, small pitting corrosion of 15 pm was observed after 1.5 years on the passive surface layer that was formed.On the basis of these results, Ti 99.8-Pd continues to be considered as a promising corrosion protection material for long-lived HLW containers.

Evaluation of the pitting corrosion of carbon steels and other ferrous metals in soil systems

Carbon steel containers are widely used in the disposal of low-level radioactive wastes in the United States. On metals in soils, the maximum pit depth due to corrosion is often expressed by h m = kt n , where t is the exposure time, k is the pitting parameter, and n is a parameter related to aeration properties of the soil. This investigation determined the dependence of k and n on soil properties based on data on carbon steels and ferrous metals obtained by the National Bureau of Standards. A linear correlation analysis of k as a function of the pH value and the resistivity revealed that k is principally influenced by the pH value of the soil. The resistivity of the soil is found to play a minor role. The following two relationships were derived: k a = 5.74(9.9 − pH) for pH < 6.8 and, k b = 5.05(2 pH − 10.3) for pH > 7.2, for acidic and alkaline soils, respectively. A linear correlation analysis of n as a function of moisture content, 0, and clay content, CL, led to the following relationship: n = A 10 + A 2( CL) + A 3. The values for A 1, A 2, and A 3 are dependent on the degree of soil aeration. Comparison of the present expressions with another available in the literature and experimental data are presented. Comparison of pitting in carbon steels and in wrought ferrous metals indicated that these materials pitted at similar rates in similar environments.

An assessment of carbon steel containers for radioactive waste disposal

The report describes a study of the corrosion of High Level Nuclear Waste containers in geological disposal. This has concentrated on the use of carbon steel in granitic environments, and has aimed at estimating the corrosion allowance required to achieve a 1000-year life. Experimental and mathematical modelling studies have indicated that 216 mm of steel should be more than sufficient to prevent penetration by general or localized corrosion. There are sound mechanistic reasons to regard this value as an overestimate. The possibility of stress corrosion cannot be dismissed, but, because the process requires a certain minimum stress level before it can occur, it should be possible to avoid the problem by giving the containers a stress relief heat treatment. A conservative assessment has shown that microbial corrosion may cause a maximum additional metal loss of 13 mm.

Predicting the Long Term Corrosion of Metal Containers for Nuclear Waste Disposal

AbstractThe prediction of the long term corrosion of metal containers for nuclear waste under geological disposal conditions requires the extrapolation of corrosion data over several hundred years. There is a general trend to use either simulation or ‘worst case’ experiments as a means of materials selection, but neither really addresses the problem of long term extrapolation. It is proposed that such an extrapolation can only be done convincingly if it is based on a sound and generally accepted mechanistic understanding of the processes involved. If such knowledge does not exist the first step must be to acquire it through experimental mechanistic studies. Subsequently such knowledge should be formulated into mathematical models, which can be used to make long term predictions, and which can be validated by comparison with short term experimental results. The application of this combined mathematical modelling/experimental approach is illustrated for three corrosion processes which may affect carbon steel containers, namely general corrosion, localised attack and stress corrosion cracking.

An Approach for Evaluating the General and Localised Corrosion of Carbon-Steel Containers for Nuclear Waste Disposal

AbstractThe paper considers the long term corrosion of carbon-steel containers for heat generating nuclear waste in a granitic repository. Under such conditions carbon steel may exhibit general, localised or passive corrosion behaviour depending on the exact composition and redox potential of the groundwater contacting the containers; localised corrosion being of most concern because it has the fastest propagation rate. It is well established, however, that such localised corrosion is only possible when the environment is sufficiently oxidising to maintain a positive potential gradient between the cathodic surface and the corrosion sites, which requires that species with oxidising potentials greater than water need to be present. This fact provides a basis for estimating the periods during which containers may be subject to localised and subsequently to general corrosion, and hence for making an overall assessment of the metal allowance required for a specified container life. A model for the diffusion transport of oxygen has been developed, and a sensitivity analysis has shown that the period of possible localised attack is strongly dependent on the passive film leakage current, the radiation dose rate and the oxygen diffusion coefficient.

TEM Investigation of Carbide Precipitation in Low Carbon Steels Containing Silicon

Etude du domaine de temps et de temperature pour la formation de carbures metastables et stables et de leur vitesse de croissance pendant le vieillissement de trois aciers a bas carbone contenant de 0,3 a 1,52% masse de silicium

種々の熱処理組織を有する炭素鋼 (0.13～0.41%C含) の疲労特性

種々の熱処理組織を有する炭素鋼 (0.13～0.41%C含) の疲労特性

TaC, TiCを含む超硬合金と炭素鋼間の拡散層について

TaC, TiCを含む超硬合金と炭素鋼間の拡散層について

A non-retreating Bunsen burner

There are two RBMK-1500 type reactors in Lithuania, which are under decommissioning now. Long-lived low and intermediate level radioactive metallic waste generated during operation and decommissioning of the nuclear power plant is foreseen for deep geological disposal (DGR). A preliminary assessment of the radionuclide release from metallic waste and migration through the engineered barriers of the generic DGR in crystalline rocks is presented in this study. The radionuclide flux to backfill and geosphere is investigated for three different container alternatives: (1) disposal of metallic waste in stainless steel containers, (2) disposal in carbon steel containers and (3) disposal in which the container is not a barrier for radionuclide migration. In addition, the impact of surface contamination of the activated waste on the radionuclide flux is investigated. Comparison of maximum fractional flux to backfill between Alternatives 1 and 2 revealed that the container material plays a noticeable role in the migration of long-lived low mobility radionuclides. The maximum fractional flux to the geosphere for all radionuclides increases from Alternative 1 to Alternative 3. Surface contamination is an important parameter when the maximum radionuclide flux to backfill is observed during the early post-closure period.