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Journal articles on the topic 'Near Surface Disposal Facility'

Author: Grafiati
Published: 6 September 2023

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1

Nazeeh, K. M., and G. L. Sivakumar Babu. "Reliability analysis of near-surface disposal facility using subset simulation." Environmental Geotechnics 6, no. 4 (June 2019): 242–49. http://dx.doi.org/10.1680/jenge.17.00004.

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2

Van Geet, M., M. De Craen, D. Mallants, I. Wemaere, L. Wouters, and W. Cool. "How to treat climate evolution in the assessment of the long-term safety of disposal facilities for radioactive waste: examples from Belgium." Climate of the Past Discussions 5, no. 1 (February 13, 2009): 463–94. http://dx.doi.org/10.5194/cpd-5-463-2009.

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Abstract. In order to protect man and the environment, long-lasting, passive solutions are needed for the different categories of radioactive waste. In Belgium, three main categories of conditioned radioactive waste (termed A, B and C) are defined by radiological and thermal power criteria. It is expected that Category A waste – low and intermediate level short-lived waste – will be disposed in a near-surface facility, whereas Category B and C wastes – high-level and other long-lived radioactive waste – will be disposed in a deep geological repository. In both cases, the long-term safety of a given disposal facility is evaluated. Different scenarios and assessment cases are developed illustrating the range of possibilities for the evolution and performance of a disposal system without trying to predict its precise behaviour. Within these scenarios, the evolution of the climate will play a major role as the time scales of the evaluation and long term climate evolution overlap. In case of a near-surface facility (Category A waste), ONDRAF/NIRAS is considering the conclusions of the IPCC, demonstrating that a global warming is nearly unavoidable. The consequences of such a global warming and the longer term evolutions on the evolution of the near-surface facility are considered. In case of a geological repository, in which much longer time frames are considered, even larger uncertainties exist in the various climate models. Therefore, the robustness of the geological disposal system towards the possible results of a spectrum of potential climate changes and their time of occurrence will be evaluated. The results of climate modelling and knowledge of past climate changes will merely be used as guidance of the extremes of climate changes to be considered and their consequences.
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3

Cho, Yeseul, Hoseog Dho, Hyungoo Kang, and Chunhyung Cho. "Evaluation of Exposure Dose and Working Hours for Near Surface Disposal Facility." Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) 20, no. 4 (December 30, 2022): 511–21. http://dx.doi.org/10.7733/jnfcwt.2022.039.

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4

Kwon, Mijin, Hyungoo Kang, and Chunhyung Cho. "Study on Rainfall Infiltration Into Vault of Near-surface Disposal Facility Based on Various Disposal Scenarios." Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) 19, no. 4 (December 30, 2021): 503–15. http://dx.doi.org/10.7733/jnfcwt.2021.042.

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5

Sucipta, Sucipta, and Suhartono Suhartono. "DETERMINATION OF CONCRETE VAULT THICKNESS OF NEAR SURFACE DISPOSAL FOR RADIOACTIVE WASTE AT SERPONG NUCLEAR AREA." Jurnal Pengembangan Energi Nuklir 19, no. 2 (April 7, 2018): 103. http://dx.doi.org/10.17146/jpen.2017.19.2.3624.

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In order to support and complement the radioactive waste management facilities in Indonesia, BATAN will build a demonstration disposal facility in Serpong Nuclear Area (SNA). Demonstration disposal that will be built is Near Surface Disposal (NSD) type. Engineered vault for NSD is reinforced concrete. The calculations for determining the thickness of NSD concrete vault is based on the conceptual design as the result of the placement optimization of demonstration disposal that takes into account the inventory of radioactive waste and environmental geology conditions of the site at Serpong Nuclear Area. The thickness of the vault in this paper is focused on its ability to withstand radiation from stored waste so that workers or people who are around the disposal facility is safe with maximum radiation dose limit rate of 0.3 μSv / h. The calculation is performed with the aid of MicroShield 7:02 and Rad Pro Calculator Version 3:26 software. From the calculation so that the dose rate at the outer surface of the vault to be 0.3 μSv / h, required walls made of concrete with a density of 2:35 g / cm3 is 62.8 cm thickness.
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6

Jang, Jiseon, Tae-Man Kim, Chun-Hyung Cho, and Dae Sung Lee. "Radiological Safety Assessment for a Near-Surface Disposal Facility Using RESRAD-ONSITE Code." Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) 19, no. 1 (March 31, 2021): 123–32. http://dx.doi.org/10.7733/jnfcwt.2021.19.1.123.

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7

Capra, B., Y. Billard, W. Wacquier, and R. Gens. "Risk assessment associated to possible concrete degradation of a near surface disposal facility." EPJ Web of Conferences 56 (2013): 05006. http://dx.doi.org/10.1051/epjconf/20135605006.

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8

Mutoni, Agnes, and Juyoul Kim. "Impact of Concrete Degradation on the Long-Term Safety of a Near-Surface Radioactive Waste Disposal Facility in Korea." Applied Sciences 12, no. 18 (September 8, 2022): 9009. http://dx.doi.org/10.3390/app12189009.

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The migration of radionuclides from radioactive waste into the environment poses a public safety concern. Thus, the long-term safety assessment for near-surface disposal sites for radioactive waste in South Korea entails providing reasonable assurance that the annual radiation dose exposure from radionuclide release from the waste repository into the biosphere will not exceed the regulatory limit of 0.1 mSv/yr. At the first near-surface disposal site in Gyeongju, concrete was a crucial component of the engineered barriers designed to contain radionuclides within the disposal site. The ability of concrete to retain radioactive waste within the disposal site is attributed to its high sorption capacity for radionuclides. However, research has shown that the degradation of concrete can affect its radionuclide retention capabilities, which are defined by sorption properties of distribution (Kd) and diffusion (Ds) coefficient parameters. As a result, changes in sorption properties may lead to radionuclides migrating out of the disposal vault. In light of the geochemical deterioration of engineered concrete barriers, this study assesses the long-term safety of near-surface disposal sites. To simulate the impact of concrete degradation on radionuclide migration, we employed RESRAD-OFFSITE’s extended source-term features, which can model the release of radionuclides from radioactive waste shielded by concrete barriers. Using carefully screened published sorption data of four radionuclides (14C, 137Cs, 90Sr and 99Tc) in different stages of concrete degradation, the results indicated that released radioactivity during the most degraded state of concrete will result in a maximum radiation exposure dose of 1.4 × 10−8 mSv/yr from 99Tc which is below the permissible limit of 0.1 mSv per year, thus demonstrating that concrete is a reliable component of the engineered designed barriers for near-surface disposal facilities.
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9

Kuzmin, E. V., A. V. Minin, M. Yu Bamborin, and Yu V. Trofimova. "System of Engineering Safety Barriers of the Facilities for Near-Surface Disposal of Radioactive Waste." Occupational Safety in Industry, no. 6 (June 2022): 46–51. http://dx.doi.org/10.24000/0409-2961-2022-6-46-51.

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Facilities for near-surface disposal of radioactive waste are very important structures consisting of several safety barriers, and the substantiation and selection of the principal structures of near-surface disposal facility is a complex task that must be solved taking into account the distinctive specifics — the time of their active and passive operation, as well as the period of potential danger of radioactive waste. The paper considers the main approaches to ensuring the long-term safety of near-surface disposal facilities through the use of various engineering safety barriers, measures to protect safety barriers, personnel, the public and the environment. To ensure safety, to prevent the spread of ionizing radiation and radioactive substances from the near-surface disposal facility into the environment, a systematization of safety barriers was carried out to ensure reliable isolation of the placed radioactive waste. Using the experience of building long-term structures, the periods of reliable isolation of the radioactive waste by each of the engineering barriers are indicated. The safety barriers, which are included as the main engineered barriers in the design solutions of the near-surface disposal facilities being created, are consistently considered. Containers are the first engineered barrier. The second barrier is a buffer material based on the natural clays that fills the space between the walls of the modular structures and containers, as well as between the containers themselves. The third barrier is concrete walls, floor slabs and floor slabs of the modular structures of the disposal site. The fourth barrier consists of bentonite mats and a clay castle made of crumpled natural clay. The fifth barrier is a multi-layered covering screen constructed for waterproofing, protection from the atmospheric precipitation, ingress of animals, plant roots and inadvertent human intrusion. The choice of materials for the engineered barriers and the requirements for the characteristics of the barrier are carried out based on the long-term safety assessment calculations, including taking into account the properties of the host rocks.
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10

Anggraini, Zeni, Jaka Rachmadetin, Nazhira Shadrina, Sucipta Sucipta, and Heru Sriwahyuni. "Modeling Radiation Exposure from Normal Release of 137Cs Radionuclide to Groundwater for Post-Closure Assessment of Serpong Near Surface Disposal Demo Facility." IOP Conference Series: Earth and Environmental Science 927, no. 1 (December 1, 2021): 012020. http://dx.doi.org/10.1088/1755-1315/927/1/012020.

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Abstract Near-surface disposal (NSD) has been applied in several countries to dispose of low-level radioactive waste. The demo plant of this disposal type is planned to be constructed in Serpong Nuclear Area, Banten. An assessment of radiation exposure is necessary to ensure the safety requirement of the facility in order to support this program. This study aims to estimate radionuclide migration from the proposed NSD demo facility to the environment and the corresponding total human dose using AMBER mathematical modeling. The representative radionuclide,137Cs, was selected because of its high mobility in the environment and the relatively long half-life in the low-level waste inventory. The scenario considered in the modeling was the normal release to the environment through groundwater. Parameters such as initial radionuclide concentration, soil physical parameters of the study site, and disposal design were entered into AMBER software to be calculated using mathematical formulas. The results show that the radionuclide concentration value in the environment is below the safe limit recommended by the Environmental Supervisory Agency. Likewise, the maximum dose received by the community around the facility is 7.40×10-11 mSv/y, 550 years after the post-closure of the facility, which is also below the regulatory limit of 1 mSv/y for the public.
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11

Suskin, V. V., I. V. Kapyrin, and F. V. Grigorev. "Assessing the efficiency of a “buried wall” barrier in the establishment of near-surface long-term storage and disposal facilities for RW." Radioactive Waste 14, no. 1 (2021): 96–105. http://dx.doi.org/10.25283/2587-9707-2021-1-96-105.

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The article evaluates the impact of a “buried wall” barrier on the long-term safety during the long-term storage1 or in-situ disposal of nuclear legacy facilities, in particular, industrial reservoirs, as well as during the development of near-surface disposal facilities for radioactive waste (RWDF). For assessment purposes, filtration and mass transfer processes have been numerically modelled in the GeRa code based on a case study of a reference near-surface facility. The study explores in which way the available covering screen affects the dynamics of contaminant spread. It evaluates the sensitivity of the results to the dispersion parameter commonly characterized by a high degree of uncertainty.
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12

Jakimavičiūtė-Maselienė, Vaidotė, Jonas Mažeika, and Stasys Motiejūnas. "Application of vadose zone approach for prediction of radionuclide transfer from near-surface disposal facility." Progress in Nuclear Energy 88 (April 2016): 53–57. http://dx.doi.org/10.1016/j.pnucene.2015.11.016.

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13

Serebryakov, B. E. "Assessment of the dose to the public from a near-surface radioactive waste disposal facility." Atomic Energy 80, no. 1 (January 1996): 53–56. http://dx.doi.org/10.1007/bf02415756.

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14

Tokarevsky, O., and I. Iarmosh. "Assessing Impact of Sorption in Geological Medium on Permissible Activity of Radioactive Waste in Near-Surface Disposal Facilities." Nuclear and Radiation Safety, no. 3(75) (August 22, 2017): 34–39. http://dx.doi.org/10.32918/nrs.2017.3(75).06.

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The paper considers the conservative scenario of potential exposure that envisages simultaneous destruction of barriers with simultaneous release of radionuclides by the example of Lot 3 near-surface radioactive waste disposal facility at the Vektor Industrial Complex located in the Chornobyl Exclusion Zone. A conceptual model that considers migration of radionuclides through the aeration zone and aquifer to the potable water well, as well as mixing of infiltration water containing radionuclides with ground water in case of reaching the aquifer, was developed to analyze the mentioned scenario. Permissible specific activity of radioactive waste in the facility is calculated under the assumption that radioactive waste contains only 90Sr radionuclide. Normalysa software is used for calculations.
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15

Anisimov, N. A., and A. A. Kuvaev. "Numerical Modeling of Moisture Transfer in the Structures of a Near-Surface Radioactive Waste Disposal Facility." Radioactive Waste 20, no. 3 (2022): 97–106. http://dx.doi.org/10.25283/2587-9707-2022-3-97-106.

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The paper considers a model used to calculate the filtration flow in the structural elements of a near-surface radioactive waste disposal facility (RWDF). It presents the results of calculations focused on the filtration and humidity field rates inside the RWDF taking into account changes in the properties of the upper screen, concrete and packages with radioactive waste occurring over time. The study also analyzes the influence of the filtration flow field features on the safety barriers state.
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16

Kim, Hyun-Joo, Minseong Kim, and Jin Beak Park. "Improvement of Safety Approach for Accidents During Operation of LILW Disposal Facility : Application for Operational Safety Assessment of the Near-surface LILW Disposal Facility in Korea." Journal of the Nuclear Fuel Cycle and Waste Technology(JNFCWT) 15, no. 2 (June 30, 2017): 161–72. http://dx.doi.org/10.7733/jnfcwt.2017.15.2.161.

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17

Yamsani, Sudheer Kumar, Sreedeep Sekharan, and Ravi R. Rakesh. "Combined Shear and Seepage Characteristics for Selecting Drainage Layer in Near Surface Hazardous Waste Disposal Facility." Geotechnical and Geological Engineering 35, no. 2 (January 10, 2017): 871–78. http://dx.doi.org/10.1007/s10706-016-0135-2.

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18

Setiawan, Budi, and Heru Sriwahyuni. "Determination of 137Cs Elimination from Solution By Tasikmalaya Bentonite and Belitung Quartz Sand As Barrier Material Candidate on the Near Surface Disposal Facility." Jurnal Kimia VALENSI 4, no. 1 (May 31, 2018): 14–21. http://dx.doi.org/10.15408/jkv.v4i1.7325.

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Batch technique experiment was applied to measure the elimination of 137Cs from solution using bentonite from Tasikmalaya and quartz sand from Belitung. Bentonite material was used as barrier system surrounding on a near surface disposal facility, and quartz sand as backfill material. The distribution coefficient (Kd) of 137Cs on bentonite and quartz sand samples have been measured. Contact time, variation of Na and Cs ion concentrations in solution were applied as the experiment parameters. The Kd values of 137Cs on the samples were 1700 and 3200 ml/g for bentonite and 17 and 37 ml/g for quartz sand samples, respectively. The Na and Cs concentrations in solution affected the Kd values of 137Cs on samples. Isotherm sorption result shown that the interaction of 137Cs onto solid samples was approached with Freundlich model. The data obtained from the experiments and then could be used for radionuclides migration assessment models of disposal facility in the future.DOI:http://dx.doi.org/10.15408/jkv.v4i1.7325
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19

Burne, S., H. S. Wheater, A. P. Butler, P. M. Johnston, P. Wadey, G. Shaw, and J. N. B. Bell. "Radionuclide Transport above a Near‐Surface Water Table: I. An Automated Lysimeter Facility for Near‐Surface Contaminant Transport Studies." Journal of Environmental Quality 23, no. 6 (November 1994): 1318–29. http://dx.doi.org/10.2134/jeq1994.00472425002300060028x.

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20

Gurumoorthy, C. "Experimental methodology to assess migration of iodide ion through Bentonite-Sand Backfill in a Near Surface Disposal Facility." Indian Journal of Science and Technology 5, no. 1 (January 20, 2012): 1–6. http://dx.doi.org/10.17485/ijst/2012/v5i1.13.

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21

Van Veelen, A., O. Preedy, J. Qi, G. T. W. Law, K. Morris, J. F. W. Mosselmans, M. P. Ryan, N. D. M. Evans, and R. A. Wogelius. "Uranium and technetium interactions with wüstite [Fe1–xO] and portlandite [Ca(OH)2] surfaces under geological disposal facility conditions." Mineralogical Magazine 78, no. 5 (October 2014): 1097–113. http://dx.doi.org/10.1180/minmag.2014.078.5.02.

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AbstractIron oxides resulting from the corrosion of large quantities of steel that are planned to be installed throughout a deep geological disposal facility (GDF) are expected to be one of the key surfaces of interest for controlling radionuclide behaviour under disposal conditions. Over the lengthy timescales associated with a GDF, the system is expected to become anoxic so that reduced Fe(II) phases will dominate. Batch experiments have therefore been completed in order to investigate how a model reduced Fe-oxide surface (wüstite, Fe1–xO) alters as a function of exposure to aqueous solutions with compositions representative of conditions expected within a GDF. Additional experiments were performed to constrain the effect that highly alkaline solutions (up to pH 13) have on the adsorption behaviour of the uranyl (UO22+) ion onto the surfaces of both wüstite and portlandite [Ca(OH)2; representative of the expected cementitious phases]. Surface co-ordination chemistry and speciation were determined by ex situ X-ray absorption spectroscopy measurements (both X-ray absorption near-edge structure analysis (XANES) and extended X-ray absorption fine structure analysis (EXAFS)). Diffraction, elemental analysis and XANES showed that the bulk solid composition and Fe oxidation state remained relatively unaltered over the time frame of these experiments (120 h), although under alkaline conditions possible surface hydroxylation is observed, due presumably to the formation of surface hydroxyl complexes. The surface morphology, however, is altered significantly with a large degree of roughening and an observed decrease in the average particle size. Reduction of U(VI) to U(IV) occurs during adsorption in almost all cases and this is interpreted to indicate that wüstite may be an effective reductant of U during surface adsorption. This work also shows that increasing the carbonate concentration in reactant solutions dramatically decreases the adsorption coefficients for U on both wüstite and portlandite, consistent with U speciation and surface reactivity determined in other studies. Finally, the EXAFS results include new details about exactly how U bonds to this metal oxide surface.
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22

Lee, Wang Hyeon, and Jae Hak Cheong. "Potential radiological hazard and options to cope with consequences from recycling of activated metal waste disposed of at a near-surface disposal facility." Annals of Nuclear Energy 152 (March 2021): 107993. http://dx.doi.org/10.1016/j.anucene.2020.107993.

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23

Rao, Sudhakar M., and P. Raghuveer Rao. "Role of the vadose zone in mitigating strontium transport at the near-surface disposal facility (NSDF) in Kalpakkam, India." Bulletin of Engineering Geology and the Environment 75, no. 4 (July 23, 2015): 1485–91. http://dx.doi.org/10.1007/s10064-015-0772-3.

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24

Arakelyan, A. A., A. I. Blohin, P. A. Blohin, Yu E. Vaneev, S. T. Kazieva, P. A. Kizub, V. G. Kondakov, S. V. Panchenko, and I. V. Sipachev. "Refinement of KORIDA software complex and its application in addressing SNF and RW management problems." Radioactive Waste 20, no. 3 (2022): 107–16. http://dx.doi.org/10.25283/2587-9707-2022-3-107-116.

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The paper overviews current progress in the development of the KORIDA software and some examples of its application providing solution to RW and SNF management problems. Its functional capacity has been greatly extended, in particular, providing automated development of models required to calculate exposure doses to personnel based on laser scanning data, as well as through the development of modules evaluating exposure impact on the population and biota. The paper also presents the results of nuclide kinetics module verification and validation, evaluated doses of potential exposure for the population in the vicinity of a near-surface RW disposal facility.
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25

Walke, R. C., M. C. Thorne, and S. Norris. "Biosphere studies supporting the disposal system safety case in the UK." Mineralogical Magazine 76, no. 8 (December 2012): 3225–32. http://dx.doi.org/10.1180/minmag.2012.076.8.35.

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AbstractHigher activity radioactive wastes remain hazardous for extremely long timescales, of up to hundreds of thousands of years. Disposing of such wastes deep underground presents the internationally accepted best solution for isolating them from the surface environment on associated timescales. Geological disposal programmes need to assess potential releases from such facilities on long timescales to inform siting and design decisions and to help build confidence that they will provide an adequate degree of safety. Assessments of geological disposal include consideration of the wastes, the engineered facility, the host geology and the surface and near-surface environment including the biosphere. This paper presents an overview of recent post-closure biosphere assessment studies undertaken in support of the Nuclear Decommissioning Authority Radioactive Waste Management Directorate disposal system safety case for geological disposal of the United Kingdom's higher activity radioactive wastes. Recent biosphere studies have included: (1) ensuring that the United Kingdom's approach to consideration of the biosphere in safety case studies continues to be fit for purpose, irrespective of which site or sites are considered in the United Kingdom's geological disposal programme; (2) updating projections of global climate and sea level, together with consideration of the potential importance of transitions between climate states; (3) considering geosphere–biosphere interface issues and their representation, including redox modelling and catchment-scale hydrological modelling; and (4) identifying key radionuclides and developing a series of reports describing their behaviour in the biosphere together with an evaluation of associated implications for post-closure assessment calculations.
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26

Hong, Sung-Wook, Sangho Park, and Jin Beak Park. "Safety Assessment on the Human Intrusion Scenarios of Near Surface Disposal Facility for Low and Very Low Level Radioactive Waste." Journal of Nuclear Fuel Cycle and Waste Technology 14, no. 1 (March 30, 2016): 79–90. http://dx.doi.org/10.7733/jnfcwt.2016.14.1.79.

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27

Sujitha, S., Deepthi Mary Dilip, and G. L. Sivakumar Babu. "Time-dependent reliability analysis for radionuclide migration in groundwater in near surface disposal facility using the enhanced Monte Carlo method." Georisk: Assessment and Management of Risk for Engineered Systems and Geohazards 11, no. 2 (September 16, 2016): 208–14. http://dx.doi.org/10.1080/17499518.2016.1229867.

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Hallam, R. J., N. D. M. Evans, and S. L. Jain. "Sorption of Tc(IV) to some geological materials with reference to radioactive waste disposal." Mineralogical Magazine 75, no. 4 (August 2011): 2439–48. http://dx.doi.org/10.1180/minmag.2011.075.4.2439.

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AbstractOne of the most important isotopes to be considered for disposal in the proposed UK Geological Disposal Facility (GDF) for higher-activity radioactive wastes will be 99Tc, due to its long half-life, high fission yield and ability to migrate through the geosphere as the pertechnetate ion. Much of the Tc is likely to be in the lower Tc(IV) oxidation state due to the low Eh in the near field. Batch Tc(IV) sorption experiments have been performed from pH 3–13, using 95mTc as a spike, in the presence of quartz, hematite, goethite, plagioclase feldspar, sand and shale. Solutions containing Tc(IV) were prepared at trace concentrations to avoid precipitation of TcO2. Values for the partition coefficient (Rd) were found to range from ∽4 up to ∽2 × 104 cm3 g–1. Rd was heavily dependent on pH in all cases, with the highest values being found in the circumneutral area. These data will inform the performance assessment for the near-field behaviour of technetium in the UK's planned higher-activity wastes GDF. Surface complexation modelling has been performed, fitting the data using both monodentate and bidentate binding models.
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Dewanto, Pandu, Setyo Sarwanto Moersidik, and Sucipta Sucipta. "Radionuclide Release Prediction in Water and Soil at Demonstration Plant of Near Surface Disposal for Radioactive Waste." Indonesian Journal of Physics and Nuclear Applications 1, no. 2 (June 30, 2016): 116. http://dx.doi.org/10.24246/ijpna.v1i2.116-122.

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Near Surface Disposal (NSD) for Radioactive Waste that should be developed due to increment of the low level radioactive waste, need to be analyzed and evaluated related to the radiological impact of the environment. A research method applied is done by modeling the distribution of radionuclide releases process. Analysis related with the releases of radionuclide in water and soil is using PRESTO (Prediction of Radiological Effects Due to Shallow Trench Operations). The application scenarios selected in this safety assessment is the migrations of Co-60 and Cs-137 scenario through the shallow groundwater flow pattern in the NSD site. The SigmaPlot software is also used to determine the concentration equation in well water and river water. The final results showed the concentration of radionuclide in wells and streams below the provision. Radionuclide activity concentrations in well ranged from 10<sup>-10</sup>Bq/m<sup>3</sup> to 10<sup>0</sup>Bq/m<sup>3</sup> and in the river ranged from 10<sup>-15</sup>Bq / m<sup>3</sup> to 10<sup>-1</sup>Bq / m3. The impact of radioactive waste of radionuclide Co-60 and Cs-137 will decrease to the background radiation level at a distance less than 10m and penetrate into the saturated layer up to 4m. In this study, an equation have been obtained that can predict radionuclide concentration patterns based on the distance and the depth of the ground surface against to the facility operation time.
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Liu, Dong-Xu, Xiao-Wei Xiong, Jin-Sheng Wang, Li-Tang Hu, and Rui Zuo. "Derivation of a speciic activity limit for plutonium for near surface disposal a case study at a potential site in northwest China." Nuclear Technology and Radiation Protection 33, no. 3 (2018): 307–16. http://dx.doi.org/10.2298/ntrp1803307l.

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Based on the safety assessment framework and site-specific characteristic investigations in northwest China, an approach to deriving the specific activity limit of 239Pu is applied to establish a proposed value. Our analyses, in conjunction with the results of other previous studies, allow for the following conclusions: (1) As an intrusion scenario with a feature of minimal site-dependence and pervasive applicability, the drilling scenario can be used as the limiting scenario for the post-closure period; (2) Given a dose limit of 5 mSv per year, a derived specific activity of 287 Bqg-1 (at a disposal depth shallower than 5 m) for 239Pu is obtained through the formulation of models and subsequent calculations. It is suggested that both our approaches to deriving the limit and the results can be effectively applied to establish acceptance criteria of long-lived transuranic nuclides, for the particular disposal facility; and (3) From the standpoint of exploring the approach for limit derivation, the intrusion scenario and the corresponding exposure evaluation can be the focus of concern in the study area. It is implied that, in arid regions, a leaching scenario may lead to a more complex evaluation, with unnecessary effort, and can be virtually excluded.
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31

Hong, Sung-Wook, Jin-Baek Park, and Jung-Hyun Yoon. "Study on the Institutional Control Period Through the Post-drilling Scenario Of Near Surface Disposal Facility for Low and Intermediate-Level Radioactive Waste." Journal of the Nuclear Fuel Cycle and Waste Technology 12, no. 1 (March 30, 2014): 59–68. http://dx.doi.org/10.7733/jnfcwt.2014.12.1.59.

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Baxter, S., D. Holton, S. Williams, and S. Thompson. "Predictions of the wetting of bentonite emplaced in a crystalline rock based on generic site characterization data." Geological Society, London, Special Publications 482, no. 1 (September 7, 2018): 285–300. http://dx.doi.org/10.1144/sp482.8.

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AbstractA geological disposal facility (GDF) is the widely accepted long-term solution for the management of higher-activity radioactive waste. It consists of an engineered facility constructed in a suitable host rock. The facility is designed to inhibit the release of radioactivity by using a system consisting of engineered and natural barriers. The engineered barriers include the wasteform, used to immobilize the waste, the waste disposal container and any buffer material used to protect the container. The natural barrier includes the rocks in which the facility is constructed. The careful design of this multi-barrier system enables the harmful effects of the radioactivity on humans and biota in the surface environment to be reduced to safe levels.Bentonite is an important buffer material used as a component of a multi-barrier disposal system. For example, compacted bentonite rings and blocks are used to protect the copper container, used for the disposal of spent fuel, in the KBS-3 disposal system. As the bentonite saturates, through contact with groundwater from the host rock, it swells and provides a low hydraulic conductivity barrier, enabling the container to be protected from deleterious processes, such as corrosion. The characteristic swelling behaviour of bentonite is due to the presence of significant quantities of sodium montmorillonite.Recently, there have been detailed in situ experiments designed to understand how bentonite performs under natural conditions. One such experiment is the Buffer–Rock Interaction Experiment (BRIE), performed at the Äspö Hard Rock Laboratory near Oskarshamn in the SE of Sweden. This experiment is designed to further understand the wetting of bentonite from the groundwater flow in a fractured granite host rock.In this paper, the observations from the BRIE are explained using an integrated model that is able to describe the saturation of bentonite emplaced in a heterogeneous fractured rock. It provides a framework to understand the key processes in both the rock and bentonite. The predictive capability of these models was investigated within the context of uncertainties in the data and the consequence for predictions of the wetting of emplaced bentonite. For example, to predict the wetting of emplaced bentonite requires an understanding of the distribution of fracture transmissivity intersecting the bentonite. A consequence of these findings is that the characterization of the fractured rock local to the bentonite is critical to understanding the subsequent wetting profiles. In particular, prediction of the time taken to achieve full saturation of bentonite using a simplified equivalent homogeneous description of the fractured host rock will tend to be too short.
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Sakib, Khondokar, Abu Haydar, Idris Ali, Debasish Paul, and Shah Alam. "Regional scale screening of selected regions of Bangladesh to identify potential sites for the disposal of low and intermediate level radioactive waste." Nuclear Technology and Radiation Protection 36, no. 1 (2021): 25–37. http://dx.doi.org/10.2298/ntrp210219010s.

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Disposal of radioactive wastes has emerged as a vital issue for Bangladesh as the country is actively working to have a nuclear power plant operating in the country by 2023-2024. The current study aims to find potential sites for a near-surface disposal facility using a geographic information system software and multi-criteria analysis method. Previously six regions (Region-1 to Region-6) were identified upon performing continental scale screening of the whole territory of Bangladesh. In the current study, regional scale screening has been performed of Region-1 and Region-2 using five criteria divided into fifteen sub-criteria: earthquakes, wind speed, rainfall, cultivated-vegetated land, forests, buildings-facilitie-built up areas (area), buildings-facilities-industries-institutions (Point), population density, medium-broad road and railway, narrow road, monument, power line, ground water table, surface water body, and lastly flooding were used in the analysis. The suitability map and relative importance weighting of these sub-criteria were determined by using the geographic information system and multi-criteria analysis method. The overlay analysis was performed of suitability maps of each sub-criterion and found the final suitability map of Region-1 and Region-2. These suitability maps were divided into six categories: the excluded area, most suitable, suitable, moderately suitable, unsuitable, and completely unsuitable. Nineteen potential sites with a maximum and minimum area of 7.90 km2 and 1.15 km2 were identified from these most suitable and suitable areas. Detailed field investigation and site characterization are needed to be performed on selected potential sites to choose a final disposal site for the low and intermediate levels of radio-active waste.
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Charles, Christopher, Simon Rout, Andrew Laws, Brian Jackson, Sally Boxall, and Paul Humphreys. "The Impact of Biofilms upon Surfaces Relevant to an Intermediate Level Radioactive Waste Geological Disposal Facility under Simulated Near-Field Conditions." Geosciences 7, no. 3 (July 12, 2017): 57. http://dx.doi.org/10.3390/geosciences7030057.

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35

Kim, Minseong, Sung-Wook Hong, and Jin Beak Park. "Uncertainty Management on Human Intrusion Scenario Assessment of the Near Surface Disposal Facility for Low and Intermediate-Level Radioactive Waste: Comparative Analysis of RESRAD and GENII." Journal of the Nuclear Fuel Cycle and Waste Technology(JNFCWT) 15, no. 4 (December 15, 2017): 369–80. http://dx.doi.org/10.7733/jnfcwt.2017.15.4.369.

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36

Kwon, Ki Nam, and Jae Hak Cheong. "Development of a reference framework to assess stylized human intrusion scenarios using GENII Version 2 considering design features of planned near-surface disposal facility in Korea." Nuclear Engineering and Technology 51, no. 6 (September 2019): 1561–74. http://dx.doi.org/10.1016/j.net.2019.04.019.

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37

Shabalin, Borys, Olena Lavrynenko, and Kostiantyn Yaroshenko. "Investigation of the insulating properties of the Cherkasy deposit clays for the creation of underlying screens of radioactive waste at the ‘Vector’ site." Proceedings of the NTUU “Igor Sikorsky KPI”. Series: Chemical engineering, ecology and resource saving, no. 2 (June 28, 2021): 71–81. http://dx.doi.org/10.20535/2617-9741.2.2021.235870.

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The lack of scientifically substantiated requirements, comprehensively developed and approved in a prescribed manner, for the usage of clays as a barrier material poses risks for the safe disposal of radioactive waste in facilities at the ‘Vector’ site for the period of their operation and closure. The bentonite clay from Ukraine’s largest Cherkasy deposit of bentonite and palygorskite clays is considered the most durable as the main component of the insulating (underlying) screens of radioactive waste disposal facilities. The main properties and compositional features of the Cherkasy natural bentonite clay (Dashukovskaya site, layer II) and its variety such as alkaline earth bentonite (activated soda bentonite), which provide isolation of radioactive waste in disposal, are considered. It is shown that the Cherkasy field has good waterproofing and barrier properties, including a high sorption capacity with respect to 90Sr and 137Cs, which is one of the main characteristics that ensure the safe disposal of radioactive waste. The alkaline earth bentonite absorbs 90Sr and 137Cs more efficiently than natural bentonite does. However, 90Sr is sorbed in larger quantities than 137Cs on both types of bentonite. With increasing time of interaction with an aqueous solution, both types demonstrate a redistribution of the mobile (exchangeable) and immobile (non-replaceable) forms of radionuclides. The contribution of the stationary form that does not participate in migration processes also increases. A comprehensive analysis of the bentonite clays of the Cherkasy deposit was carried out, taking into account the significance of recoverable reserves and the potential for improving the technical and economic parameters of clays. Thus, the Cherkasy bentonite clays can be recommended as an additional anti-migration engineering barrier for ground/near-surface facilities for the disposal of radioactive waste. When choosing the type of bentonite clay for use as a barrier in a radioactive waste disposal facility, one could take into account the data published in the article, but the question of applying the bentonite clays of the Cherkasy deposit to ensure the safe disposal of radioactive waste remains to be further studied.
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38

Lee, Christopher A., Arjen van Veelen, Katherine Morris, J. Fred W. Mosselmans, Roy A. Wogelius, and Neil A. Burton. "Uranium (VI) Adsorbate Structures on Portlandite [Ca(OH)2] Type Surfaces Determined by Computational Modelling and X-ray Absorption Spectroscopy." Minerals 11, no. 11 (November 8, 2021): 1241. http://dx.doi.org/10.3390/min11111241.

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Portlandite [Ca(OH)2] is a potentially dominant solid phase in the high pH fluids expected within the cementitious engineered barriers of Geological Disposal Facilities (GDF). This study combined X-ray Absorption Spectroscopy with computational modelling in order to provide atomic-scale data which improves our understanding of how a critically important radionuclide (U) will be adsorbed onto this phase under conditions relevant to a GDF environment. Such data are fundamental for predicting radionuclide mass transfer. Surface coordination chemistry and speciation of uranium with portlandite [Ca(OH)2] under alkaline groundwater conditions (ca. pH 12) were determined by both in situ and ex situ grazing incidence extended X-ray absorption fine structure analysis (EXAFS) and by computational modelling at the atomic level. Free energies of sorption of aqueous uranyl hydroxides, [UO2(OH)n]2–n (n = 0–5) with the (001), (100) and (203) or (101) surfaces of portlandite are predicted from the potential of mean force using classical molecular umbrella sampling simulation methods and the structural interactions are further explored using fully periodic density functional theory computations. Although uranyl is predicted to only weakly adsorb to the (001) and (100) clean surfaces, there should be significantly stronger interactions with the (203/101) surface or at hydroxyl vacancies, both prevalent under groundwater conditions. The uranyl surface complex is typically found to include four equatorially coordinated hydroxyl ligands, forming an inner-sphere sorbate by direct interaction of a uranyl oxygen with surface calcium ions in both the (001) and (203/101) cases. In contrast, on the (100) surface, uranyl is sorbed with its axis more parallel to the surface plane. The EXAFS data are largely consistent with a surface structural layer or film similar to calcium uranate, but also show distinct uranyl characteristics, with the uranyl ion exhibiting the classic dioxygenyl oxygens at 1.8 Å and between four and five equatorial oxygen atoms at distances between 2.28 and 2.35 Å from the central U absorber. These experimental data are wholly consistent with the adsorbate configuration predicted by the computational models. These findings suggest that, under the strongly alkaline conditions of a cementitious backfill engineered barrier, there would be significant uptake of uranyl by portlandite to inhibit the mobility of U(VI) from the near field of a geological disposal facility.
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39

Leterme, B., D. Mallants, and D. Jacques. "Estimation of future groundwater recharge using climatic analogues and Hydrus-1D." Hydrology and Earth System Sciences Discussions 9, no. 1 (January 30, 2012): 1389–410. http://dx.doi.org/10.5194/hessd-9-1389-2012.

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Abstract. The impact of climate change on groundwater recharge is simulated using climatic analogue stations, i.e. stations presently under climatic conditions corresponding to a given climate state. The study was conducted in the context of a safety assessment of a future near-surface disposal facility for low and intermediate level short-lived radioactive waste in Belgium; this includes estimating groundwater recharge for the next millennia. Groundwater recharge was simulated using the Richard's based soil water balance model Hydrus-1D and meteorological time series from analogue stations. Water balance calculations showed that transition from a temperate oceanic to a warmer subtropical climate without rainfall seasonality is expected to yield a decrease in groundwater recharge (−12% for the chosen representative analogue station of Gijon, Northern Spain). Based on a time series of 24 yr of daily climate data, the long-term average annual recharge decreased from 314 to 276 mm, although total rainfall was higher (947 mm) in the warmer climate compared to the current temperate climate (899 mm). This is due to a higher soil evaporation (233 mm versus 206 mm) and higher plant transpiration (350 versus 285 mm) under the warmer climate.
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40

Shabalin, B. H., К. К. Yaroshenko, and S. P. Buhera. "Peculiarities of 137Cs Sorption/Desorption by Bentonite Clays of Cherkasy Deposit from Groundwater Model Solutions of Radioactive Waste Disposal Facilities at the “Vector” Production Complex." Nuclear Power and the Environment 21, no. 2 (2021): 78–87. http://dx.doi.org/10.31717/2311-8253.21.2.8.

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The main feature of bentonite clays is their high sorption capacity with respect to various radionuclides. The study of sorption kinetics of 137Cs was performed in the static mode by natural and industrial soda modified (PBA-20) samples of bentonite clays of Cherkasy deposit of bentonite and paligorskite clays from groundwater model solutions of radioactive waste disposal facilities of “Vector” production complex under various pH and solution mineralisation. The desorption of occluded samples was studied in distilled water and acetateammonium buffer solution. The value of the degree of sorption (S) for 137Cs on the modified samples exceeds 90%, for natural bentonite this indicator is lower (about 83–85%). On both types of bentonite with increasing time of their contact with aqueous solution and pH, there is a redistribution of water-soluble, ion-exchange and fixed forms of radionuclide and the share of the latter, that is not participating in migration processes increases, indicating the ability of bentonites to immobilize effectively for a long time. It is shown that Na-modified bentonite has higher proportion of sorption in fixed form compared to natural one and its application increases the probability of irreversible fixation of migrating radionuclides under non-optimal conditions of sorption (high pH (>11) of water after prolonged contact with cement-concrete components of engineering barriers) and thus increases the environmental safety of the storage facility. It is shown that bentonite clays of the Cherkasy deposit can serve as an effective material for creating anti-migration barriers of I and II stages of surface/near-surface storage facilities for radioactive waste disposal at the “Vector” production complex. At the same time, the issue of practical application of bentonite clays of Cherkasy deposit for accurate predictions of securing radioactive waste disposal of Chornobyl origin requires further study of sorption-desorption properties of bentonite clay with respect to other fission products and actinides
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41

Panchal, Y., I. M. Mohamed, Dale Pierce, N. Mounir, O. Abou-Sayed, Loloi, and Abou-Sayed. "An Economic, Technical and Environmental Feasibility Study for Slurry Injection for Biosolids Management in the Dallas Fort Worth Metroplex." Journal of Solid Waste Technology and Management 46, no. 1 (February 1, 2020): 24–35. http://dx.doi.org/10.5276/jswtm/2020.24.

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In Dallas Fort Worth (DFW), sewage is treated with a combination of anaerobic digestion, effluent filtration and lime stabilization to create biosolids which are then composted, landfilled, or land applied. The current treatment procedure has certain concerns including emissions or accumulation of odors, pathogens, nutrients, metals, and pharmaceutical products.<br/> An alternative method, the Slurry Injection technique, enables the digestion of biosolids in the deep earth and can replace the current practice of wastewater treatment or disposal in a much more environmentally friendly and cost-efficient manner. By completely sequestering methane and CO2 into deep geologic formations which are produced as biosolids breakdown, reduces the greenhouse gas emissions and enables the operator to create greenhouse gas emission offset credits which can be marketed to offset the operating costs.<br/> The economic, environmental, and technical aspects of building a new biosolids slurry injection facility in DFW, includes both the surface construction requirements as well as the subsurface strata evaluation for containment assurance. For the subsurface aspects, a geomechanical and stress analysis is performed on the Atoka formation (near the city of Fort Worth) and it confirms a confining layer above and below the injection zone to keep the waste contained for permanent storage.
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42

Shaikh, Janarul, Sudheer Kumar Yamsani, Sanjeet Sahoo, Sreedeep Sekharan, and Ravi Ranjan Rakesh. "Hydraulic performance assessment of a multi-layered landfill cover system under constant water ponding." Acta Horticulturae et Regiotecturae 25, no. 2 (November 1, 2022): 129–40. http://dx.doi.org/10.2478/ahr-2022-0017.

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Abstract The engineered multi-layered cover system (MLCS) is used to minimize rainwater infiltration into the wastes accommodated in near surface waste disposal facility (NSDF). It is important to assess the hydraulic performance of MLCS before deploying it in the field. For this purpose, an instrumented three-layered soil column representing MLCS was subjected to 1.5 m constant ponding head for 400 days. The variation of volumetric water content and soil water potential was monitored as a function of depth and time. The objective of the study is to understand the long-term hydraulic performance and rate of saturation of different layers of MLCS. Under constant water ponding, the time to saturation for 0.3 m in surface layer, 0.6 m in drainage layer and 1.0 m in hydraulic barrier layer was observed as 24, 223 and 262 days, respectively. The numerical analysis of the MLCS predicted comparable time duration of 25, 234 and 272 days, respectively. It was noted that the numerical simulation performed by using measured wetting hydraulic parameters matched well with the experimental observation. The importance of soil specific calibration of water content sensors to improve the accuracy of observations was demonstrated. Percentage error in the estimation of layer specific soil water storage, clearly indicates that the volumetric water content measurements using profile probe was marginally better than 5TM measurements.
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43

Baston, G. M. N., M. M. Cowper, T. G. Heath, T. A. Marshall, and S. W. Swanton. "The effect of cellulose degradation products on thorium sorption onto hematite: studies of a model ternary system." Mineralogical Magazine 76, no. 8 (December 2012): 3381–90. http://dx.doi.org/10.1180/minmag.2012.076.8.51.

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AbstractCellulose degradation products (CDPs) can complex with radioelements causing solubility enhancement and sorption reduction, effects which are detrimental to the containment of radionuclides in the near field of a geological disposal facility and surrounding geosphere. Isosaccharinic acid (ISA) is the principal component of CDPs formed under the alkaline anaerobic conditions of a cement-based near field and appears to be largely responsible for the impact of CDPs on radionuclide solubility and sorption under near-field conditions. However, the situation appears to be more complicated under near-neutral pH geosphere conditions.A combined experimental and modelling study was undertaken to compare the impact of a CDP leachate to ISA in a simple model ternary sorption system consisting of hematite as a single mineral substrate, thorium as the radioelement and ISA or a CDP leachate as the complexant. Thorium sorbs strongly to hematite. A triple layer model for thorium sorption to hematite was refined to fit to the experimental data in the absence of ISA or CDP leachate; the effect of ISA on thorium sorption was then predicted.In the presence of CDP leachate, a significant reduction in thorium sorption was observed from pH 6 to 12 in good agreement with model predictions based on a high concentration of ISA. However, only a limited impact of ISA on thorium sorption was observed at pH 6 to 12, in contrast to predictions. The effects of ISA could be accounted for by assuming the formation of a ternary thorium–ISA–surface complex. The model has yet to be extended to the more complex CDP systems. Differences in the thorium speciation in solution due to the formation of a ternary calcium–thorium–ISA complex in the CDP leachate, which is absent from solutions with ISA only, provides the most likely explanation for the differences observed experimentally.
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44

Leterme, B., D. Mallants, and D. Jacques. "Sensitivity of groundwater recharge using climatic analogues and HYDRUS-1D." Hydrology and Earth System Sciences 16, no. 8 (August 6, 2012): 2485–97. http://dx.doi.org/10.5194/hess-16-2485-2012.

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Abstract. The sensitivity of groundwater recharge to different climate conditions was simulated using the approach of climatic analogue stations, i.e. stations presently experiencing climatic conditions corresponding to a possible future climate state. The study was conducted in the context of a safety assessment of a future near-surface disposal facility for low and intermediate level short-lived radioactive waste in Belgium; this includes estimation of groundwater recharge for the next millennia. Groundwater recharge was simulated using the Richards based soil water balance model HYDRUS-1D and meteorological time series from analogue stations. This study used four analogue stations for a warmer subtropical climate with changes of average annual precipitation and potential evapotranspiration from −42% to +5% and from +8% to +82%, respectively, compared to the present-day climate. Resulting water balance calculations yielded a change in groundwater recharge ranging from a decrease of 72% to an increase of 3% for the four different analogue stations. The Gijon analogue station (Northern Spain), considered as the most representative for the near future climate state in the study area, shows an increase of 3% of groundwater recharge for a 5% increase of annual precipitation. Calculations for a colder (tundra) climate showed a change in groundwater recharge ranging from a decrease of 97% to an increase of 32% for four different analogue stations, with an annual precipitation change from −69% to −14% compared to the present-day climate.
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45

Bain, Daniel J., Tetiana Cantlay, Brittany Garman, and John F. Stolz. "Oil and gas wastewater as road treatment: radioactive material exposure implications at the residential lot and block scale." Environmental Research Communications 3, no. 11 (November 1, 2021): 115008. http://dx.doi.org/10.1088/2515-7620/ac35be.

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Abstract The resurgence of oil and gas extraction in the Appalachian Basin has resulted in an excess of oil and gas brines in Pennsylvania, West Virginia, and Ohio. Primarily driven by unconventional development, this expansion has also impacted conventional wells and consequently, created economic pressure to develop effective and cheap disposal options. Using brine as a road treatment, directly or as a processed deicer, however, creates substantial concern that naturally occurring radioactive material in the brines can contaminate roads and road-side areas. Current decision making is based on risk exposure scenarios developed by regulatory agencies based on recreational users in rural areas and exposures to drivers during a typical commute. These scenarios are not appropriate for evaluating exposures to residential deicer users or people living near treated streets. More appropriate exposure scenarios were developed in this work and exposures predicted with these models based on laboratory measurements and literature data. Exposure scenarios currently used for regulatory assessment of brine road treatment result in predicted exposures of 0.4–0.6 mrem/year. Residential exposures predicted by the scenarios developed in this work are 4.6 mrem/year. If the maximum range of near-road soil radium concentrations observed in the region is used in this residential scenario (60 pCi/g 226Ra, 50 pCi/g 228Ra), residents living near these roads would be exposed to an estimated 296 mrems/year, above regulatory exposure thresholds used in nuclear facility siting assessments. These results underline the urgent need to clarify exposure risks from the use of oil and gas brines as a road treatment, particularly given the existing disparities in the distribution of road impacts across socioeconomic status.
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46

Wissmeier, Laurin, and Joachim Poppei. "Simulating the feedback between corrosive gas generation and water availability for the evaluation of radionuclide mobility in the context of radioactive waste disposal." Safety of Nuclear Waste Disposal 1 (November 10, 2021): 109–10. http://dx.doi.org/10.5194/sand-1-109-2021.

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Abstract. It has been recently recognized that the availability of liquid water may be a controlling factor in the feedback between the physical processes of variably saturated liquid and gas flow on the one hand, and various chemical processes such as metal corrosion in an underground storage facility for radioactive waste on the other hand (e.g., Huang et al., 2021, and reference therein). Iron corrosion in anoxic conditions produces hydrogen gas and consumes water, as expressed by the following stylized chemical equation (e.g., Diercks and Kassner, 1988; Senior et al., 2021): 3Fe+4H2O⟶Fe3O4+4H2 Since water is an educt the corrosion reaction may be suspended or suppressed by the scarcity of water near the corroding surfaces. At the same time, gas pressure build-up through hydrogen generation may limit further water ingress. We developed a model that focuses on the close coupling between gas generation through iron corrosion and water availability. The feedback between iron corrosion, gas generation and liquid phase flow is considered by implementing the corrosion reaction in the subsurface flow and transport simulator PFLOTRAN (Hammond et al., 2012; Lichtner et al., 2015, 2020) making use of its coding provisions to implement source/sink terms for water and gas. These source/sink terms reflect the kinetics of the iron corrosion and its dependence on the educts, where the availability of water is approximated by the local liquid saturation. The model was applied to evaluate the mobility of radionuclides in, and their release from a hypothetical geological storage facility for radioactive waste. The radionuclides are traced through the emplacement chambers and drift by means of advective and diffusive transport. Parameter variations illustrate the influence of crucial modelling parameters on the simulation results.
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de Visser-Týnová, Eva, Stephen W. Swanton, Stephen J. Williams, Marcel P. Stijkel, Alison J. Walker, and Robert L. Otlet. "14C release from irradiated stainless steel." Radiocarbon 60, no. 6 (November 22, 2018): 1671–81. http://dx.doi.org/10.1017/rdc.2018.134.

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ABSTRACTRadiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.
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48

Hutchinson, D. E., and L. F. Toussaint. "Near-surface disposal of concentrated NORM wastes." Applied Radiation and Isotopes 49, no. 3 (March 1998): 265–71. http://dx.doi.org/10.1016/s0969-8043(97)00247-9.

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49

Kim, Ki Beom, Jang Hwa Lee, and Do Gyeum Kim. "Microstructure Analysis on LILW Waste Disposal Facility by Accelerated Steel Corrosion Tests." Applied Mechanics and Materials 378 (August 2013): 194–97. http://dx.doi.org/10.4028/www.scientific.net/amm.378.194.

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Concrete structures such as LILW waste disposal facility located near the sea may suffer from chloride attack damages. This study aims to analysis mock-up test for acceleration corrosion of reinforcing bar and its deterioration in concrete structures by XRD and chloride diffusion coefficient. Corrosion acceleration experiment test has been developed and used to evaluate the effect of corrosion of reinforcing bar caused by seawater on engineered barrier of LILW waste disposal facility.
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50

Durkee, John. "Managing a Surface Cleaning Facility." Metal Finishing 110, no. 6 (July 2012): 32–33. http://dx.doi.org/10.1016/s0026-0576(13)70218-6.

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