atw 2017-09

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atw Vol. 62 (2017) | Issue 8/9 ı August/September 542 DECOMMISSIONING AND WASTE MANAGEMENT Acronyms and Abbreviations ATLAS Advanced Thermal-Hydraulic Test Loop for Accident Simulation DBA Design Basis Accident ECCS Emergency Core Cooling Systems KAERI Korea Atomic Energy Research Institute LOCA Loss of Coolant Accident Corrosion of Canister Materials for Radioactive Waste Disposal Bernhard Kienzler In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe (today Karlsruhe Institute of Technology, KIT). The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Boundary conditions were defined by the expected temperatures and salt brines relevant under accidental conditions. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in “grey literature” which is hardly accessible any more. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel. Introduction Since 1980, research and development activities were carried out at the (Nuclear) Research Center Karlsruhe (today KIT) on the corrosion of materials which might be suitable for construction of containers for highly radioactive waste. The data reported in this contribution relate almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). At this time, it was assumed that the radioactive wastes could be vitrified and disposed of after relatively short periods of time. The boundary conditions for the research on container materials for highly radioactive waste resulted from the requirements defined by pouring the molten highly radioactive glass directly into the canister, the necessary welding and decontamination of the containers and, on the other hand, by the requirement for transport, interim storage and final disposal. The first repository concepts considered simple containers. Only when the direct disposal of used nuclear fuels was considered since about 1994, overpacks or multi-shell containers were regarded seriously. The POLLUX container [1], which consisted of an MSSV Main Steam Safety Valve POSRV Pilot-Operated Safety Relief Valve RCP Reactor Coolant Pump SIP Safety Injection Pump SIT Safety Injection Tank SG Steam Generation SPACE Safety & Performance Analysis Code for nuclear power plants TLOFW Total Loss of Feedwater inner container capable of absorbing the rock pressure in a salt formation and an outer container shielding the radioactive radiation, was developed for the burned nuclear fuel. The investigations were performed until 2004 and covered a selection of steels and other metals, as well as ceramic materials. Most of the experiments were performed under closed conditions in autoclaves and the samples were immersed in various salt Material Authors Minhee Kim Seyun Kim Central Research Institute Korea Hydro & Nuclear Power Co. 70 1312-gil Yuseong-daero, Yuseong-gu, Daejeon, 305-343, Korea brines. Temperatures between 35 °C and 200 °C were applied in order to simulate the conditions for low-level and high-level radioactive waste disposal in a rock salt mine. In parallel to the laboratory tests a series of in-situ corrosion tests have been conducted. Most of these tests were performed in the Asse II salt mine aiming on high temperature corrosion data. The most important outcome of the corrosion experiments are Material description Material number Density g/cm 3 Ni based alloys Hastelloy C4 Ni Mo 16 Cr 16 Ti 2.4610 8.669 Ti alloys Titan – Palladium Ti 99.7 – Pd 3.7025 4.593 Ti 99.7 - Pd EG 3.7035 4.593 Fe based alloys Fine-grained steel FStE 255 1.0566 7.814 TStE 460 1.8915 7.671 15 Mn Ni 6.3 1.6210 7.512 DC 01 / St 12 1.0330 7.85 ST 37-2 1.0038 7.856 Nodular cast steel GGG 40.3 0.7043 6.955 Cr-Ni steel 1.4833 8.022 Cu alloys Cu.99 4.0000 9.198 Cu-Ni 70/30 4.7000 8.866 Cu-Ni 90/10 4.9000 8.998 Ni alloys Nickel 99.9 2.4068 8.48 Ni/Cu 70/30 2.4360 8.51 | | Tab. 1. Metal alloys for construction of waste canisters under investigation. Decommissioning and Waste Management Corrosion of Canister Materials for Radioactive Waste Disposal ı Bernhard Kienzler
atw Vol. 62 (2017) | Issue 8/9 ı August/September summarized recently [2]. Table 1 presents the list of all metal alloys under investigations which were considered to be suited for constructing nuclear waste canisters. This paper presents a summary of results for fine-grained steels and nodular cast steel obtained in corrosion studies performed at KIT between 1980 and 2004. Results of experiments under irradiation or in the presence of trace components, e.g. sulfide or hydrogen, contact corrosion studies, hydrogen embrittlement in the case of Ti-Pd and of fine-grained steels as well as the effect of different welding procedures are not discussed in the present contribution. Due to the recommendations of the commission [3] and the German law for amendment of the site selection law (passed by the German Parliament, March 23, 2017 [4]) some boundary conditions have been changed. These conditions refer not only to the potential host rocks, but also to the maximum temperature at the canister surfaces which is now limited to 100 °C, and the demand for retrieveability of the wastes during the operational phase of the disposal and the recoverability of the wastes for a period of 500 years. Materials and Methods Corrosion experiments with steels made use of hot rolled and normalized sheet, the nodular cast steel was investigated in the cast conditions. The tests were mainly performed by immersion of samples in different corrosion media, such as saturated NaCl- and MgCl 2 -richt solutions. To establish realistic environmental conditions for a deep repository, the tests were conducted in autoclaves, preventing the presence of oxygen and CO 2 from air. Under the pH MgCl 2 rich solution 1 (mol / l) conditions of the salt solutions (see Table 2), fine-grained steels and nodular cast steel show general or uniform corrosion. At the expected pH values of the solutions, the formation of a passivation layer is not possible [5]. General corrosion results in a relatively uniform mass loss over the entire area of the sample. Its effects predictable. corrode uniformly when exposed to open atmospheres, soils and natural waters, leading to the rusty appearance. As long as diffusion of dissolved oxygen is controlling corrosion of cast irons and carbon steels, moderate Cl- and SO 4 2- ions concentrations have essentially no effect on the overall corrosion rate of steel because they do not affect the solubility of oxygen. Above a critical pH, passivation occurs, and the corrosion rate of the passivated steel increases with Cl - concentration [5]. Localized attack, however, is accelerated as the conductivity of water increases with increase in the concentration of these ions, expanding the effective cathodic areas. The samples were immersed in the salt brines for several months (up to 13 yrs.) at temperatures between 35 °C and 200 °C. The composition of the salt solutions discussed in this contribution are listed in Tab. 2. After termination of the experiment, the samples were removed from the test vessels and analyzed with respect to mass loss, pitting corrosion, crevice corrosion and stress-corrosion- cracking. Corrosion products were analyzed by XRD and the pH and Eh of the solutions were determined. Besides the laboratory investigations, in-situ corrosion experiments were conducted in the Asse II salt mine and were integrated in several European research programs. The in-situ experiments in rock salt based MgCl 2 rich solution 2 (mol / l) NaCl solution (Solution 3) (mol / l) | | Tab. 2. Composition of the saturated salt solutions used for the immersion tests [6, 7]. Konrad water (mol / l) Na + 0.31 0.06 4.66 2.81 K + 0.78 0.02 Mg 2+ 3.43 4.17 0.02 0.10 Ca 2+ 0.24 0.02 0.34 Cl - 7.63 8.90 4.66 3.67 SO 2 -4 0.16 0.03 0.01 pH (electrode) 5.5 5.7 5.9 6.3 -log m H+ 7.4 8.0 6.6 6.9 Density (g / ml) 1.30 1.33 1.20 Ionic strength (mol/l) 11.5 13.3 4.8 4.1 | | Fig. 1. Recovery of corrosion samples after termination of an in-situ experiment in a drill hole in the Asse II mine. After termination of the experiment, a new gallery was mined to get access to the heater and samples. The white ring was the sample holder. on the concept of high level waste disposal in vertical drill holes or in horizontal galleries. The results are documented in a series of reports [8-11]. In the drill hole tests, the samples were arranged around heaters and recovered after termination of the experiments. Figure 1 shows the recovery action for one of these experiments. In-situ experiments aiming on low level waste disposal were also performed. For this purpose several storage conditions for drums were simulated [12]: • Drums were stored in rock salt consisting of 95 wt. % NaCl and 5 wt. % CaSO 4 and were covered with a salt layer. The average water content of the rock salt is about 0.1 wt. % [13]. • Storage of free standing drums in a tunnel in-the dry atmosphere of the Asse mine at relative air humidity of 22%. • Storage in salt brine with a composition of solution 1 (Tab. 2). The measured pH value was 5.5 and the oxygen content measured at 25 °C was 0.15 mg/l. For all samples, the storage temperature was approx. 32 °C which corresponds to the temperature prevailing on the 750 m level of the Asse salt mine. [12]. Also in the context of the investigations for Konrad iron ore mine, corrosion studies have been performed, both in laboratory and under in-situ conditions. Corrosion behavior of canister steels samples (St 12, ST 37-2) and cast nodular steel (GGG 40.3) were analyzed. DECOMMISSIONING AND WASTE MANAGEMENT 543 Decommissioning and Waste Management Corrosion of Canister Materials for Radioactive Waste Disposal ı Bernhard Kienzler
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