• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2356-2365
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• 2023

Characterization, dismantling and pre-disposal management of irradiated graphite (i-graphite) have an important role in safe decommissioning of several nuclear facilities which used this material as moderator and reflector. In addition to common radiation protection issues, easily volatizing long-lived radionuclides and stored Wigner energy could be released during imprudent retrieval and processing of i-graphite. With this regard, among all cutting technologies, abrasive waterjet (AWJ) can successfully achieve all of the thermo-mechanical and radiation protection objectives. In this work, factorial experiments were designed and systematically conducted to characterize the AWJ processing parameters and the machining capability. Moreover, the limitation of dust production and secondary waste generation has been addressed since they are important aspects for radiation protection and radioactive waste management. The promising results obtained on non-irradiated nuclear graphite blocks demonstrate the applicability of AWJ as a valid technology for optimizing the retrieval, storage, and disposal of such radioactive waste. These activities would benefit from the points of view of safety, management, and costs.

• Nuclear Engineering and Technology
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• v.56 no.11
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• pp.4820-4825
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• 2024

Scientific literature studies irradiated graphite treatment. Research also covers graphite conditioning and its long-term behavior under disposal conditions. The European Commission's CARBOWASTE project, titled "Treatment and disposal of irradiated graphite and other carbonaceous waste", is a key reference for state-of-the-art studies on alternative solutions. It identified 24 strategic options for managing irradiated graphite throughout its complete life cycle. The methodology proposed in this paper entails the application of the Analytic Hierarchy Process (AHP) method to rank the 24 options, placing particular emphasis on the weighting of seven criteria for selecting management options for the irradiated graphite. The highest weights were assigned by experts to 'environment and public safety' (28.05 %) and 'worker safety' (26.16 %). The objective is to develop a standardized approach enabling waste management companies to identify the most appropriate management option, considering structural and legislative constraints in their operating country. Examining the study findings, option 19 "In-situ entombment" stands out as the best choice in both the CARBOWASTE project and the proposed methodology. Thus, this methodology could assist hypothetical entities in examining management options for irradiated graphite, with the aim of identifying the optimal solution for graphite waste disposal.

• Structural Engineering and Mechanics
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• v.50 no.6
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• pp.723-740
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• 2014

A fracture criterion based on the strain energy density (SED) over a control volume, which embraces the notch edge, is employed in the present paper to assess the fracture loads of some U-notched Brazilian disk (UNBD) specimens. The specimens are made of commercial graphite and have been tested under pure mode I, pure mode II and mixed mode I/II loading. The results show that the SED criterion allows to successfully assess the fracture loads of graphite specimens for different notch tip radii and various mode mixity conditions with discrepancies that fall inside the scatter band of ${\pm}20%$.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.211-218
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• 2013

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ($^{14}C$), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the $^{14}C$, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create $CO_x$ gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide ($CO_2$). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2002.07a
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• pp.25-37
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• 2002

The most important industrial application of gamma radiation in characterizing green compacts is the determination of the density. Examples are given where this method is applied in manufacturing technical components in powder metallurgy. The requirements imposed by modern quality management systems and operation by the workforce in industrial production are described. The accuracy of measurement achieved with this method is demonstrated and a comparison is given with other test methods to measure the density. The advantages and limitations of gamma ray densitometry are outlined. The gamma ray densitometer measures the attenuation of gamma radiation penetrating the test parts (Fig. 1). As the capability of compacts to absorb this type of radiation depends on their density, the attenuation of gamma radiation can serve as a measure of the density. The volume of the part being tested is defined by the size of the aperture screeniing out the radiation. It is a channel with the cross section of the aperture whose length is the height of the test part. The intensity of the radiation identified by the detector is the quantity used to determine the material density. Gamma ray densitometry can equally be performed on green compacts as well as on sintered components. Neither special preparation of test parts nor skilled personnel is required to perform the measurement; neither liquids nor other harmful substances are involved. When parts are exhibiting local density variations, which is normally the case in powder compaction, sectional densities can be determined in different parts of the sample without cutting it into pieces. The test is non-destructive, i.e. the parts can still be used after the measurement and do not have to be scrapped. The measurement is controlled by a special PC based software. All results are available for further processing by in-house quality documentation and supervision of measurements. Tool setting for multi-level components can be much improved by using this test method. When a densitometer is installed on the press shop floor, it can be operated by the tool setter himself. Then he can return to the press and immediately implement the corrections. Transfer of sample parts to the lab for density testing can be eliminated and results for the correction of tool settings are more readily available. This helps to reduce the time required for tool setting and clearly improves the productivity of powder presses. The range of materials where this method can be successfully applied covers almost the entire periodic system of the elements. It reaches from the light elements such as graphite via light metals (AI, Mg, Li, Ti) and their alloys, ceramics ($AI_20_3$, SiC, Si_3N_4, $Zr0_2$, ...), magnetic materials (hard and soft ferrites, AlNiCo, Nd-Fe-B, ...), metals including iron and alloy steels, Cu, Ni and Co based alloys to refractory and heavy metals (W, Mo, ...) as well as hardmetals. The gamma radiation required for the measurement is generated by radioactive sources which are produced by nuclear technology. These nuclear materials are safely encapsulated in stainless steel capsules so that no radioactive material can escape from the protective shielding container. The gamma ray densitometer is subject to the strict regulations for the use of radioactive materials. The radiation shield is so effective that there is no elevation of the natural radiation level outside the instrument. Personal dosimetry by the operating personnel is not required. Even in case of malfunction, loss of power and incorrect operation, the escape of gamma radiation from the instrument is positively prevented.