Relevant bibliographies by topics / Near Surface Disposal Facility

Academic literature on the topic 'Near Surface Disposal Facility'

Author: Grafiati
Published: 6 September 2023

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Journal articles on the topic "Near Surface Disposal Facility"

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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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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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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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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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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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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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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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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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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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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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Dissertations / Theses on the topic "Near Surface Disposal Facility"

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Smith, Donna Lee. ""Redox pumping" in the near surface Missoula aquifer iin the flood plain of the Clark Fork River surface, water and groundwater interaction and arsenic related chemistry at a compost facility near a wastewater treatment plant /." CONNECT TO THIS TITLE ONLINE, 2008. http://etd.lib.umt.edu/theses/available/etd-06062008-105818/.

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Van, den Berg Mader J. "Western limb tailings reclamation project." Diss., Pretoria : [s.n.], 2004. http://upetd.up.ac.za/thesis/available/etd-07272005-100249.

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Raghuveer, Rao P. "Factors Influencing Contaminant Transport in Vadose Zone of Near Surface Radioactive Waste Disposal Facility." Thesis, 2015. https://etd.iisc.ac.in/handle/2005/4802.

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Radioactive waste is generally classified as low and intermediate level waste (LILW) and high-level waste (HLW), based on their level of radioactivity and the time taken for decay. The low and intermediate level solid/solidified wastes are emplaced in near surface shallow land repository and are termed as Near Surface Disposal Facility (NSDF). A near surface disposal facility (NSDF) is proposed to be built at Kalpakkam, Tamil Nadu. Kalpakkam is situated about 70 km south of Chennai (12o33’ N Lat and 80o11’ E Long) along the east coast of India. Several of India’s nuclear installations like Madras Atomic Power Station, Fast Breeder Test Reactor, Kalpakkam Reprocessing Plant and Indira Gandhi Centre for Atomic Research (IGCAR) are located in Kalpakkam. Typical reinforced concrete trench (RCT) for disposal of low and intermediate level nuclear waste have major portion of RCT being located below the ground surface and above the water table, i.e., in the vadose zone. Available studies on the mineralogical and physico-chemical characterization of clays at Kalpakkam nuclear plant site have not focused on the mineralogical, physico-chemical and hydraulic properties of soils in vadose zone. The physico-chemical properties of soils are important as they strongly affect the fate and mobility of radioactive contaminants in the sub-surface soil. The hydraulic properties of soils in the vadose zone differ from those in the saturated zone owing to dis-continuity in water filled voids that makes the flow path more tortuous; additionally presence of suction in soil voids renders gradient causing flow to vary with the soil moisture content. Strontium is the selected contaminant solute in this thesis as the waste inventory of proposed NSDF at Kalpakkam shall predominantly contain radioactive strontium and cesium ions. The unsaturated (vadose) zone housing the buried portion of RCT is important as hydraulic characteristics (void ratio, degree of saturation, volumetric water content, suction and unsaturated permeability coefficient) influence the transport of moisture and thereby of the solute dissolved in the moisture to underlying groundwater table, upon breach of the RCT. The physico-chemical properties (pore water chemistry, cation exchange capacity of soils) of the vadose zone soil play significant role in solute transport by way of adsorption/ion-exchange and desorption of the contaminant. Given the significance of hydraulic and physico-chemical properties of the vadose zone soil in influencing solute transport and the lack of studies in this direction in the Indian context, the present thesis examined strontium adsorption and transport for soils obtained from the vadose zone region of proposed NSDF at Kalpakkam, Tamil Nadu, India from 2 pits (termed Pit 1 and Pit 2). Strontium adsorption studies by the NSDF soils were accomplished from batch and miscible displacement column experiments.
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Books on the topic "Near Surface Disposal Facility"

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Schulz, R. K. Control of water infiltration into near surface LLW disposal units. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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Desimone, Leslie A. Ground-water quality near a septage-treatment facility, Orleans, Massachusetts, 1997-98. Northborough, Mass: U.S. Dept. of the Interior, U.S. Geological Survey, 2000.

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R, Dayal, Han K. W, and International Atomic Energy Agency, eds. Surveillance and monitoring of near surface disposal facilities for radioactive waste. Vienna, Austria: IAEA, 2004.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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Scorca, Michael P. Ground-water quality near a scavenger-waste-disposal facility in Manorville, Suffolk County, New York 1984-85. Syosset, N.Y: Dept. of the Interior, U.S. Geological Survey, 1990.

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C, Torres-Vidal, Batandjieva B, and International Atomic Energy Agency. Waste Safety Section., eds. Safety assessment methodologies for near surface disposal facilities: Results of a co-ordinated research project. Vienna: IAEA, 2004.

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Book chapters on the topic "Near Surface Disposal Facility"

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Rumynin, Vyacheslav G. "Radon Site for Near-Surface Disposal of Solid RW." In Theory and Applications of Transport in Porous Media, 617–45. Dordrecht: Springer Netherlands, 2011. http://dx.doi.org/10.1007/978-94-007-1306-2_21.

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Sunny, Faby, and Manish Chopra. "Radiological Impact Assessment for Near Surface Disposal of Thorium Waste." In Thorium—Energy for the Future, 205–13. Singapore: Springer Singapore, 2019. http://dx.doi.org/10.1007/978-981-13-2658-5_14.

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Jianqin, Liu, Xiong Xiaowei, Sun Qinghong, and An Hongxiang. "Derivation of Activity Limits of Spent Radioactive Source for Near-Surface Disposal." In Proceedings of The 20th Pacific Basin Nuclear Conference, 441–51. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2314-9_39.

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Liu, Jianqin, Chao Gao, Li Zhang, and Xiang Qin. "Derivation of Activity Concentration Upper Limits for Low Level Solid Radioactive Waste." In Springer Proceedings in Physics, 536–48. Singapore: Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_47.

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AbstractThe derivation method of activity concentration upper limits for low level radioactive waste is put forward. The activity concentration upper limits for low level solid radioactive waste will be ascertained, on the basis of safety assessment of near surface disposal. On the premise of basic safety requirements about radioactive waste near surface disposal, taking Yaotian disposal site and Feifengshan disposal site as the reference sites, the drilling water scenario, drilling scenario, post-drilling scenario, and housing scenario after the institutional control period when the disposal sites have been closed are considered. The radionuclide transfer process and exposure pathway of various scenarios are analyzed, the conceptual model and mathematical model of radionuclide transfer are established, and the effective dose to human from various scenarios is calculated. Assuming a linear relationship between the activity concentration and the dose, the activity concentration upper limits of various scenarios are then derived for each radionuclide that meet the appropriate dose criteria. The smallest upper limit is chosen, by the approximate integer principle, the magnitude of upper limit of each radionuclide for low level solid waste is then ascertained.
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Yu, Hong-Ying, Dong Zhao, Xin Shang, Fang Wang, and Li He. "Life-Cycle Cost Study for a Near-Surface Disposal Repository of Low-Level Waste in China." In Springer Proceedings in Physics, 733–41. Singapore: Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-19-8899-8_71.

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Rumynin, Vyacheslav G. "Tomsk-7 and Krasnoyarsk-26 Sites for Deep-Well Injection Radioactive Waste Disposal, and Lake Karachai Site of Near-Surface Disposal of Radioactive Brine." In Theory and Applications of Transport in Porous Media, 681–711. Dordrecht: Springer Netherlands, 2011. http://dx.doi.org/10.1007/978-94-007-1306-2_23.

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Tammemagi, Hans. "Are There Better Disposal Methods?" In The Waste Crisis. Oxford University Press, 2000. http://dx.doi.org/10.1093/oso/9780195128987.003.0011.

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Most of the solid waste generated by society ultimately winds up in near-surface landfills. Let us put our thinking caps firmly on, place our prejudices aside, and explore what other methods might be used to dispose of waste. We should seek, in particular, the approaches that best fulfill the three basic principles described in chapter 2. That is, we should strive to find disposal methods that are in accord with sustainable development. Existing and abandoned pits, quarries, and mines are attractive for waste disposal because a hole to contain the wastes has already been excavated. Such abandoned areas, when left unreclaimed, cannot be used for agriculture or other beneficial uses. Thus, they generally do not have significant market value and can often be obtained relatively cheaply. For these reasons, pits and quarries have been extensively used for landfills. Operating and abandoned mines, on which this section focuses, are somewhat similar to pits and quarries, though usually larger. Abandoned mines hold promise as disposal facilities because they are resource areas that have been depleted and thus have little future value. There are two basic types of mine: the open pit mine, which is effectively a large pit or hole in the ground; and the underground mine, where the mined-out openings are deep underground and there is no surface expression except for the shafts used to gain subsurface access. Because underground mines occupy minimal surface land, their use for waste disposal would be in accordance with the sustainable development principles that were advocated in chapter 2. Several European countries, with higher population densities and much smaller land mass than in North America, have long used abandoned underground mines to dispose of their rubbish. The major advantage of placing wastes deep in underground mines is that it is inherently safer than placing the wastes in a surface facility. The amount of groundwater and its flow rate decrease with depth; this fact, combined with the long transport paths back to the biosphere, minimizes the possibility that contaminants will be carried by groundwater to the surface, where they could damage the environment. The waste is contained deeper and more securely.
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Kozak, Matthew W. "Safety assessment for near-surface disposal of low and intermediate level wastes." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 475–98. Elsevier, 2017. http://dx.doi.org/10.1016/b978-0-08-100642-9.00016-5.

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Kozak, M. W. "Safety assessment for near-surface disposal of low- and intermediate-level radioactive waste." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 522–46. Elsevier, 2010. http://dx.doi.org/10.1533/9781845699789.4.522.

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Crossland, I. G. "Near-surface, intermediate depth and borehole disposal of low-level and short-lived intermediate-level radioactive waste." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 43–81. Elsevier, 2010. http://dx.doi.org/10.1533/9781845699789.1.43.

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Conference papers on the topic "Near Surface Disposal Facility"

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Pinkston, Karen, David W. Esh, and Christopher J. Grossman. "Performance Assessment for Depleted Uranium Disposal in a Near-Surface Disposal Facility." In 2008 MRS Fall Meetin. Materials Research Society, 2008. http://dx.doi.org/10.1557/proc-1124-q11-06.

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Gurumoorthy, C., and O. Kusakabe. "Centrifuge Modeling of Radioactive Waste Migration Through Backfill in a Near Surface Disposal Facility." In The 11th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2007. http://dx.doi.org/10.1115/icem2007-7032.

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Investigations on the performance of backfill barrier in Near Surface Disposal Facility (NSDF) for radioactive wastes are important to ensure the long term safety of such disposal option. Favorable condition to delay migration of radionuclides from disposed waste to far fields is diffusion process. However, advective dispersion/diffusion mechanism plays an important role due to changes in backfill over a period of time. In order to understand these mechanisms, detailed laboratory experiments are usually conducted for developing mathematical models to assess the behaviour of backfill. However, these experiments are time consuming and suffer with the limitations due to material complexity. Also, there are constraints associated with validation of theoretical predictions due to intricacy of boundary conditions as well as the time scale is quite different as compared to the time required for completion of the processes in the field. Keeping in view these aspects, centrifuge modeling technique has been adopted by various researchers to model and understand various geoenvironment problems in order to provide a link between the real life situation termed as the ‘Prototype’ and its model, which is exposed to a higher gravitational field. An attempt has been made in this paper to investigate the feasibility of this technique to model advective dispersion/diffusion mechanism of radionuclides through saturated Bentonite-Sand (B:S) backfill. Various stages of centrifuge modeling are highlighted. Column tests were conducted in the centrifuge to evaluate the hydraulic conductivity of B:S mixture under prototype NSDF stress conditions. Results showed that steady state hydraulic conductivity under saturated conditions was 2.86 × 10−11m/sec. Studies indicate the feasibility of centrifuge modeling technique and usefulness to model advective diffusion of radionuclides through B:S backfill.
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Ormai, Péter, Károly Bérci, Ferenc Takáts, and László Juhász. "Long Lived Radiation Sources and Waste in Near Surface Repository: Legacy of the Past, Challenge of Today." In ASME 2001 8th International Conference on Radioactive Waste Management and Environmental Remediation. American Society of Mechanical Engineers, 2001. http://dx.doi.org/10.1115/icem2001-1161.

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Abstract In the near surface repository highly active sources and spent sealed radiation sources (SSRS) consisting of long half-life and alpha-emitting materials have been disposed of. Based on the results of a safety assessment it is clear that the SSRS could result in high doses to individuals who intrude into the facility and high doses following any future disruption of the facility by natural processes. Intervention alternatives that are being considered cover the whole range of options from extension of the facility to terminate operation. The possible future scenarios to improve the overall safety of the site and to extend the existing disposal capacity are discussed.
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Dreimanis, A. "Enlargement of the Baldone Near-Surface Radioactive Waste Repository." In The 11th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2007. http://dx.doi.org/10.1115/icem2007-7165.

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A unified analysis of the enlargement of the Baldone near-surface radioactive waste (RW) repository RADONS considers the interplay of the existing engineering, safety and infrastructure premises, with the foreseen newly socio-technical features. This enlargement consists in construction of two additional RW disposal vaults and in building a long-term storage facility for spent sealed sources at the RADONS territory. Our approach is based on consecutive analysis of following basic elements: - the origin of enlargement – the RADONS safety analysis and a set of optimal socio-technical solutions of Salaspils research reactor decommissioning waste management; - the enlargement – a keystone of the national RW management concept, including the long-term approach; - the enlargement concept – the result of international co-operation and obligations; - arrangement optimization of new disposal and storage space; - environmental impact assessment for the repository enlargement – the update of socio-technical studies. The study of the public opinion revealed: negative attitude to repository enlargement is caused mainly due to missing information on radiation level and on the RADONS previous operations. These results indicate: basic measures to improve the public attitude to repository enlargement: the safety upgrade, public education and compensation mechanisms. A detailed stakeholders engagement and public education plan is elaborated.
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Perko, Janez, Suresh C. Seetharam, Diederik Jacques, Dirk Mallants, Wim Cool, and Elise Vermariën. "Influence of Cracks in Cementitious Engineered Barriers in a Near-Surface Disposal System: Assessment Analysis of the Belgian Case." In ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96226.

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In large cement-based structures such as a near surface disposal facility for radioactive waste voids and cracks are inevitable. However, the pattern and nature of cracks are very difficult to predict reliably. Cracks facilitate preferential water flow through the facility because their saturated hydraulic conductivity is generally higher than the conductivity of the cementitious matrix. Moreover, sorption within the crack is expected to be lower than in the matrix and hence cracks in engineered barriers can act as a bypass for radionuclides. Consequently, understanding the effects of crack characteristics on contaminant fluxes from the facility is of utmost importance in a safety assessment. In this paper we numerically studied radionuclide leaching from a crack-containing cementitious containment system. First, the effect of cracks on radionuclide fluxes is assessed for a single repository component which contains a radionuclide source (i.e. conditioned radwaste). These analyses reveal the influence of cracks on radionuclide release from the source. The second set of calculations deals with the safety assessment results for the planned near-surface disposal facility for low-level radioactive waste in Dessel (Belgium); our focus is on the analysis of total system behaviour in regards to release of radionuclide fluxes from the facility. Simulation results are interpreted through a complementary safety indicator (radiotoxicity flux). We discuss the possible consequences from different scenarios of cracks and voids.
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Hanusˇi´k, V., Z. Kusovska´, J. Bala´zˇ, and A. Mrsˇkova´. "Approach to Derivation of Waste Acceptance Criteria for Mochovce Disposal Facility." In ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4797.

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In Slovakia, low and intermediate level radioactive wastes are disposed in a near-surface repository at Mochovce site. The repository, which was commissioned in September 2001, has a disposal capacity 22,320 m3. It is a vault-type concrete structure repository with reinforced concrete containers as the final waste packages. The Mochovce repository is designed to receive acceptable radioactive wastes from decommissioned A-1 power plant at Jaslovske´ Bohunice, operational waste from NPPs V-1 and V-2 at Jaslovske´ Bohunice site and NPP Mochovce, as well as institutional radioactive wastes. Generally, calculation endpoint of disposal facilities performance assessment is radiological impact on humans and environment. In that case, starting points of assessment are the waste activity concentrations and inventory activity. The acceptance of radioactive waste in Mochovce repository is one of the many elements that directly contribute to the safety of the disposal system. In Mochovce repository safety analysis, end points are both the concentration per package and total activity values. On the other hand, radiological protection criteria are the starting points of the calculation. This approach was developed and applied because the actual inventory that will be disposed of is highly uncertain. As a result of the accidents, the primary circuit was contaminated by fission products. Some auxiliary circuits and facilities were also contaminated. The complicated problem is the relatively high content of long-lived radionuclides (inclusive transuranic elements) in some waste streams. After two technological incidents at NPP A-1 uncertainties in waste inventory are large because of variability in the types of waste streams and variability in the quality and completeness of the waste characterization data. This paper presents the philosophy of safety analysis, development of scenarios, their modelling and approach that have been used to derive waste acceptance criteria, specifically limits of activity. The approach consists of the determination of radionuclides important for safety, the use of relevant safety scenarios, the setting of dose limits associated with scenarios, the calculation of activity limits and application of the simple summation rule. Finally, information is provided about short operation of the repository.
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Poskas, P., J. E. Adomaitis, and R. Kilda. "Management of Institutional Radioactive Waste in Lithuania." In ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation. ASMEDC, 2003. http://dx.doi.org/10.1115/icem2003-4877.

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The growing number of radionuclide applications in Lithuania is mirrored by increasing demands for efficient management of the associated radioactive waste. For the effective control of radioactive sources a national authorization system based on the international requirements and recommendations was introduced, which also includes keeping and maintaining the State Register of Sources of lonising Radiation and Occupational Exposure. The principal aim of the Lithuania’s Radioactive Waste Management Agency is to manage and dispose all radioactive waste transferred to it. Radioactive waste generated during the use of sources in non-power applications are managed according to the basic radioactive waste management principles and requirements set out in the Lithuanian legislation and regulations. The spent sealed sources and other institutional waste are transported to the storage facilities at Ignalina NPP. About 35,000 spent sealed sources in about 500 packages are expected until year 2010 at Ignalina NPP storage facilities. The existing disposal facility for radioactive waste from research, medicine and industry at Maisiagala was built in the early 1960’s according to a concept typical of those applied in the former Soviet Union at that time. SKB (Sweden) with participation of Lithuanian Energy Institute has performed assessment of the long-term safety of the existing facility. It was shown that the existing facility does not provide safe long-term storage of the waste already disposed in the facility. Two alternatives were defined to remedy the situation. A first alternative is the construction of a surface barrier and a second one is a retrieval solution, whereby the already stored waste will be retrieved for conditioning, characterisation and interim storage at Ignalina NPP. Facilities for the processing of the institutional radioactive waste are required before submittal to Ignalina NPP for storage, since the present facilities are inadequate. Feasibility study to establish a new central facility has been performed by SKB International Consultants (Sweden) with participation of Lithuanian Energy Institute. This study has identified the process applied and equipment needed for a new facility. Reference design and Preliminary Safety Assessment have also been performed. Plans for the interim storage and disposal of the institutional waste are described in the paper. The aspects of finging safe disposal solutions for spent sealed sources in a near surface repositories are also discussed.
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Maranville, Victoria M., and Richard McGrath. "A Summary of Radiological Waste Disposal Practices in the United States and the United Kingdom." In ASME 2009 12th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2009. http://dx.doi.org/10.1115/icem2009-16379.

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A systematic review of near-surface repositories for radioactive waste in the United States (US) was conducted. The main focus of the review consisted of a literature search of available documents and other published sources on low level radioactive waste (LLRW) disposal practices, remediation of LLRW sites in the US, and public participation for remediation efforts of near-surface radiological waste disposal sites in the US. This review was undertaken to provide background information in support of work by the United Kingdom’s (UK) Low Level Waste Repository (LLWR) and to aid in optimizing the future management of this site. The review contained a summary of the US and UK radiological waste classification requirements including a discussion of the waste types, disposal requirements, and the differences between US and UK disposal practices. A regulatory overview and evolution of regulatory requirements in the US is presented. The UK regulatory environment is also discussed and contrasted to the US process. The public participation, as part of the US regulatory process, is provided and the mechanism for stakeholder identification and involvement is detailed. To demonstrate how remediation of radiologically impacted sites is implemented in the US, existing US case studies, in which remediation activities were carried out, were reviewed. The following information was compiled: type of wastes disposed of to US shallow ground facilities [with comparison with UK classifications], facility designs (with special emphasis on those directly comparable to the subsurface conditions in the UK), and deficiencies identified in operation or in demonstrating safe post closure; and processes and difficulties in remedial actions encountered at the selected sites. Stakeholder involvement is discussed within the case studies. Publicly available information related to radiological waste management and disposal practices were reviewed. Two sites are presented in this publication for discussion. These US sites were selected based on the site similarities to conditions in the UK.
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Pa´nik, Michal, Matej Zachar, and Vladimi´r Necˇas. "Dose Assessment of Personnel Handling Conditioned Radioactive Waste." In ASME 2009 12th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2009. http://dx.doi.org/10.1115/icem2009-16149.

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Radioactive waste may arise throughout the lifetime of any type of nuclear facility. This waste has to be isolated from the environment using the engineered and natural barriers of near surface or deep geological radioactive waste repository. Before final disposal, the waste volume is reduced in the treatment process and then it is immobilized into the stable matrix. In Slovakia, the treated radioactive waste is conditioned into fibre-reinforced concrete containers using a cementation technology. These containers are the only overpacks approved for near surface disposal in the National Radioactive Waste Repository located at Mochovce nuclearsite (Slovakia). Every filled container has to fulfill the defined limits and conditions for safe transport and disposal. The dose parameters, determining the safety of personnel handling a container, are one of them. Basically, the doses are measured on the container walls’ outer surfaces and on the lid of the container. The dose monitoring in 1 meter distance from the container walls and from the lid of the container is performed before transportation as well. The calculation code VISIPLAN 3D ALARA is a planning tool and it calculates dose parameters also for the above mentioned positions concerning the fibre-reinforced concrete container which contains waste with different physical or radiological characteristics. In the paper, calculated data are compared with in-situ measurements. Using VISIPLAN 3D ALARA planning tool, various scenarios are evaluated. Finally, the optimization leading to the lowest radiation exposure of personnel handling the filled fibre-reinforced containers is discussed.
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Jung, Hagen G., and Gabriele Bandt. "Regulations for the Disposal of Radioactive Waste in the Konrad Repository." In ASME 2011 14th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2011. http://dx.doi.org/10.1115/icem2011-59105.

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In Germany low / medium level waste, which is classified here as radioactive waste with negligible heat generation, will be disposed of in the Konrad underground repository. The construction and the operation of this nuclear facility required authorization by different fields of law, i.e., by nuclear law, mining law and water law. Whereas the nuclear law considers solely radiological aspects, the relevant permit issued according to the water law considers the impact of radioactive as well as non-radioactive harmful substances. The Federal Office for Radiation Protection (BfS) as operator of the repository and permit holder has (a) to record the disposed of radioactive and non-radioactive harmful substances and (b) to balance them. To meet these requirements BfS has developed a concept, which led to a site specific solution. Threshold values were defined for recording and for balancing the harmful substances. It had to be verified that by disposal of radioactive waste packages according to these values an adverse effect on the near-surface groundwater can be excluded. The Lower Saxony Water Management, Coastal Protection and Nature Conservation Agency (NLWKN) as the responsible water law regulatory authority approved the operator’s concept as appropriate to comply with the requirements of the Water Law Permit. Nonetheless, collateral clauses were imposed to assure this.
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Reports on the topic "Near Surface Disposal Facility"

1

Kostelnik, K. M. Performance assessments for near-surface low-level radioactive waste disposal facilities. Office of Scientific and Technical Information (OSTI), December 1991. http://dx.doi.org/10.2172/183894.

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Meyer, Philip D., Prasad Saripalli, and Vicky L. Freedman. Near-Field Hydrology Data Package for the Integrated Disposal Facility 2005 Performance Assessment. Office of Scientific and Technical Information (OSTI), June 2004. http://dx.doi.org/10.2172/15010637.

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Brandt, C. A., W. H. Jr Rickard, M. G. Hefty, and N. A. Cadoret. Interim reclamation report, Basalt Waste Isolation Project Near Surface Test Facility 1990. Office of Scientific and Technical Information (OSTI), January 1991. http://dx.doi.org/10.2172/6280059.

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4

Chipman, N. Potential near-surface disposal concepts for high-level radioactive waste at the Idaho Chemical Processing Plant. Office of Scientific and Technical Information (OSTI), December 1989. http://dx.doi.org/10.2172/6915068.

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Aaberg, R. L., and W. E. Jr Kennedy. Definition of intrusion scenarios and example concentration ranges for the disposal of near-surface waste at the Hanford Site. Office of Scientific and Technical Information (OSTI), October 1990. http://dx.doi.org/10.2172/6349821.

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Schulz, R. K., R. W. Ridky, and E. O`Donnell. Control of water infiltration into near surface LLW disposal units: Progress report on field experiments at a humid region site, Beltsville, Maryland. Office of Scientific and Technical Information (OSTI), August 1996. http://dx.doi.org/10.2172/366500.

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Schulz, R. K., R. W. Ridky, and E. O`Donnell. Control of water infiltration into near surface LLW disposal units. Progress report on field experiments at a humid region site, Beltsville, Maryland: Volume 8. Office of Scientific and Technical Information (OSTI), April 1995. http://dx.doi.org/10.2172/46691.

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Schulz, R. K., R. W. Ridky, and E. O`Donnell. Control of water infiltration into near surface low-level waste disposal units. Final report on field experiments at a humid region site, Beltsville, Maryland. Office of Scientific and Technical Information (OSTI), September 1997. http://dx.doi.org/10.2172/554206.

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9

Thorne, D. J., L. M. McDowell-Boyer, D. C. Kocher, C. A. Little, and E. K. Roemer. ORNL results for Test Case 1 of the International Atomic Energy Agency`s research program on the safety assessment of Near-Surface Radioactive Waste Disposal Facilities. Office of Scientific and Technical Information (OSTI), July 1993. http://dx.doi.org/10.2172/10167328.

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10

Wilkins, Justin, Andrew McQueen, and Burton Suedel. Improving spatial and temporal monitoring of dredging operations incorporating unmanned technologies. Engineer Research and Development Center (U.S.), August 2023. http://dx.doi.org/10.21079/11681/47520.

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The US Army Corps of Engineers (USACE) is responsible for maintaining safe and navigable waterways through the periodic dredging of shoaled sediment from federal navigation channels. While dredging, a portion of the bottom sediments become resuspended creating a sediment plume near the dredging operation. Suspension of sediments during dredging and dredged sediment disposal operations continues to be a primary concern of regulatory agencies charged with the protection of environmental resources. Consequently, almost all dredging projects incorporate some level of regulatory compliance monitoring dedicated to measuring sediment resuspension. For numerous reasons the conventional approach using manned surface vessels to perform compliance monitoring is frequently ineffective in both adaptively managing dredging projects and ensuring true environmental protection. Advancements in unmanned platforms and payload technologies offer new and potentially more robust alternatives to conventional platforms. In this study, the use of unmanned aerial system (UAS) and weather balloon mounted camera imagery was demonstrated, and the use of an unmanned surface vessel (USV) to monitor turbidity in navigation channels and near a dredging operation. The imagery from the UAS and weather balloon were compared to in-situ turbidity measurements in a turbid distributary channel and near a dredging operation, while the USV was used to learn more about in-situ turbidity associated with passing vessels in a navigation channel. The results of the demonstrations show the unmanned technology bundled with off-the-shelf payloads can help to produce evidence-based information through easily interpreted aerial imagery and in situ measurements which can help to inform and manage water quality in areas where sediment plumes are an environmental concern.
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