Diffusion of cesium and iodine in compressed IG-110 graphite compacts

Carter, Lukas M.; Brockman, John; Robertson, John; Loyalka, S.K.

doi:10.1016/j.jnucmat.2016.04.024

Cited by 13 publications

(5 citation statements)

References 9 publications

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“…On the other hand, IG-110 graphite with an average pore size of 2 µm has become an ideal candidate material for reactors due to its higher graphitization degree, high temperature stability, strength and stiffness, low thermal expansion coefficient, and corrosion resistance. Therefore, Carter [5,6] tested the diffusion coefficients of I − and I 2 in bulk IG-110 graphite and compacted IG-110 graphite using ICP-MS, and the results showed that there was no significant difference in iodine diffusion coefficients between bulk IG-110 and compacted IG-110 graphite, both at the order of 10-10. They further described the low sensitivity of iodine diffusion to graphite pore size.…”

Section: Introductionmentioning

confidence: 99%

Diffusion Behavior of Iodine in the Micro/Nano-Porous Graphite for Nuclear Reactor at High Temperature

Qi,

Lian,

et al. 2023

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The diffusion behavior of iodine in micro/nano-porous graphite under high-temperature conditions was studied using analysis methods such as Rutherford backscattering Spectrometry, scanning electron microscopy, X-ray diffraction, and Raman spectroscopy. The results indicate that iodine diffusion leads to the Lattice Contractions in Microcrystals, a decrease in interlayer spacing, and a rise of defect density. And the reversal or repair of microstructure change was observed: the microcrystal size of the graphite increases, the interlayer spacing appears to return to the initial state, and the defect density decreases, upon diffusion of iodine out of iodine-loaded graphite. The comparative study comparing the iodine diffusion performance of nanoporous graphite (G400 and G450) between microporous graphite (G500), showed that nanoporous graphite exhibits a better barrier to the iodine diffusion. The study on the diffusion behavior of iodine in micro/nano-porous graphite holds substantial academic and engineering value for the screening, design, and performance optimization of nuclear graphite.

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Section: Introductionmentioning

confidence: 99%

Diffusion Behavior of Iodine in the Micro/Nano-Porous Graphite for Nuclear Reactor at High Temperature

Qi,

Lian,

et al. 2023

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“…Wenzel et al [6] determined the C-14 in the graphite matrix of spent AVR fuel element and agreed with the C-14 profile in a spent AVR sphere as published in a previous literature [7] (see Figure 1). Some other literatures [8][9][10] have introduced the progress of fundamental studies on the behaviors of H-3, Cs, or I in the HTGR system. Besides, some research results on irradiated graphite material used in British Magnox reactors are also useful as [11][12][13].…”

Section: Introductionmentioning

confidence: 99%

Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

Liu¹,

Huang²,

Xie³

et al. 2017

Science and Technology of Nuclear Installations

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The high temperature gas-cooled reactor (HTGR) has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR) is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10) in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

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“…The time-release method described in this work has previously been successfully applied to diffusion measurements of Cs, I, Ag, and most recently Sr in several types of nuclear graphite [51,[58][59][60][61][62][63]. The Arrhenius parameters of Sr were Ea = 346 (± 22) kJ/mol and D0 = 1.7 (± 1.3) × 10 begin to contribute to diffusion processes resulting in increased rates of migration [66].…”

Section: Discussionmentioning

confidence: 99%

“…The time-release experiment described in this work has previously been successfully applied to effective diffusion measurements of Cs, I, Ag, and Sr in several types of graphite [51,[58][59][60][61][62][63][64]. The experiment assumes that migration and release of the Eu diffusant from the graphite is dominated by diffusion, and that effects from adsorption and trapping are minimal [52].…”

Section: Theorymentioning

confidence: 99%

“…The time-release method used here has been described in previous publications and is only briefly described here [51,[58][59][60][61][62][63][64]. In Once experiments were complete, INAA was performed on the samples in the same manner as previously described in order to determine the total mass loss of Eu over 102 the course of each experiment.…”

Section: Diffusion Measurementsmentioning

confidence: 99%

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Fission product diffusion in ig-110 graphite

Weilert¹

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The threat of global climate change and resultant disasters has never been higher. The promises made by many countries of carbon neutrality by 2050 will be impossible to achieve without nuclear technology. Global public support for nuclear energy is at its highest level in modern history but is still severely hampered by perceptions of safety and issues involving nuclear waste and proliferation. The Fukushima disaster of 2011 and the attack on Ukraine by the Russian Federation in 2022 only served to highlight the dangers of older reactor technology and the potential for large releases of radioactivity either by accident or intentional sabotage. For these reasons most countries have a keen interest in improved reactor technologies, particularly in regards to safety, as they plan to build, or continue building, their nuclear fleets. Generation-IV reactors are characterized by improved safety, economics, and proliferation resistance compared to current light water reactor designs. The hightemperature gas-cooled reactor (HTGR) exemplifies these characteristics with the additional benefit of process heat production capabilities. Past and current demonstration reactors have proved the technical feasibility of the design and several future reactors are set to enter demonstration phases as early as the late 2020s. Despite strong performance in past and present reactors, there remains several unknown variables, particularly in regards to fission product behavior and transport under differing reactor conditions. Due to the robust nature of the tristructural isotropic fuel particles used in HTGRs, as well as the large amount of graphite comprising the core, there is little risk of a reactor meltdown. Instead, the primary safety consideration of HTGRs is the release of radioactive materials from the core, either during normal operation or an off-normal event. Most fission and activation products will be completely retained in either the fuel particle or the surrounding matrix graphite; a few, however, have a demonstrated ability to migrate through all core structures and deposit onto cooling system components. This poses a danger to reactor workers and, if the closed coolant circuit were to be compromised, the public. With that in mind, it is essential to fully understand the transport parameters of these select radionuclides in every component of the reactor core, including the core structural graphite. This work has measured effective diffusion coefficients of Sr, Ag, Pd, Eu, and Cs in IG-110 structural graphite. A time-release method was utilized to measure these diffusion coefficients at temperatures up to 1973 K using an inductively-coupled plasma mass spectrometer. The effective diffusion coefficients here reported can be used to aid predictive fission product transport programs.

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Diffusion of cesium and iodine in compressed IG-110 graphite compacts

Cited by 13 publications

References 9 publications

Diffusion Behavior of Iodine in the Micro/Nano-Porous Graphite for Nuclear Reactor at High Temperature

Diffusion Behavior of Iodine in the Micro/Nano-Porous Graphite for Nuclear Reactor at High Temperature

Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

Fission product diffusion in ig-110 graphite

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