Safe Decommissioning Research Articles

Detection of deposits adhered on the back surface of plate-like structures using a scanning laser source technique

An imaging technique using a scanning laser source (SLS) was applied to detect deposits inside pipes, which are necessary for the safe decommissioning of the Fukushima Nuclear Power Plant. Experimental results showed that the more firmly adhered the deposit on the back surface of a flat plate, the clearer it is to obtain an image of the deposit. This result is as expected because the imaging by an SLS technique is dependent on the bending stiffness of the thin plate structure. Furthermore, even epoxy putty as large as 50 mm in diameter adhered to the inner surface of the pipe could be imaged, and even when the receiver device was changed from a piezoelectric device to a non-contact laser doppler vibrometer, the image of the deposit could be obtained properly with some degradation of the image due to the effect of a lower signal-to-noise ratio.

Uptake and coordination behaviour of uranyl on functionalised silica

The safe decommissioning of nuclear reactors is a complex process, requiring a range of technological options for it to be done effectively. It has become more important in recent years, with the substantial number of reactors entering the decommissioning phase. This paper details the use of functionalised silica for the recovery of uranium from aqueous hydrochloric acid solutions, a potential agent for the treatment of steel reactor components during nuclear reactor decommissioning. Silica functionalised with either phosphonic acid, bistriazine, or bistriazinylbipyridine ligands was shown to extract uranium, with the commercially available phosphonic acid functionality being the most effective. Extractions of 100 % were observed at [H+] ≤ 0.24 M, further the phosphonic acid functionality also had the highest loading capacity. Increasing [H+] has a suppressive effect on uranyl recovery, which was more pronounced for the in-house synthesized silica, and was attributed to protonation of the extracting moiety. Extended x-ray absorption fine structure spectroscopy was also used to determine the uranium coordination environments on the functionalised silica, and gain insights into the extraction mechanism. The uranyl cation was bound by two phosphonic acid groups, two bis-triazine groups, and one bistriazinylbipyridine group, respectively on each respective extractant. Uranyl was found to be penta-coordinate in the equatorial plane for all systems, with oxygen and/or chloride always present in the first coordination sphere. Evaluation of the mechanism indicated that uranyl must be loading onto the phosphonic acid functionality via multiple mechanisms, to both the phosphonic acid and silica related surface groups, although preferential binding of uranyl to the phosphonic acid groups was observed.

Guideline for the Decommissioning/Abandonment of Subsea Pipelines

The operating lifespan of pipelines is limited, defined by their specific design codes and specifications, with the economic justification for this being determined primarily by the pipeline owner. During its operational lifespan, a pipeline’s integrity is affected mainly by the quality of the hydrocarbons being transported. The integrity of a pipeline can be maintained with regular inspections and maintenance/cleaning programmes followed from installation to commissioning. As production matures and declines, operators face several decisions concerning the pipeline’s future. There are several potential scenarios, and each should be assessed on a case-by-case basis for any specific pipeline in question. The industry best practices outline the minimum requirements for the safe decommissioning of pipelines. However, there currently need to be international specifications to be followed for the decommissioning of disused offshore pipelines. This paper aims to provide insight into the decommissioning and abandonment of offshore pipelines. Also, this article provides case studies for the decommissioning of subsea pipelines.

Dissolution behavior and aging of iron–uranium oxide

Understanding the dissolution behavior of the fuel debris is necessary for the safe decommissioning of Fukushima Daiichi Nuclear Power Plants. The dissolution behavior of FeUO4 compounds formed by a high-temperature reaction of UO2 with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using U3O8 and Fe3O4 as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy–energy-dispersive X-ray spectroscopy, and X-ray absorption fine-structure spectroscopy. The fission products were produced via thermal-neutron irradiation. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO4 compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO4 dissolution. It was thermodynamically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free U3O8).

Decommissioning: another critical challenge for energy transitions

To achieve the dual goals of minimising global pollution and meeting diverse demands for environmental justice, energy transitions need to involve not only a shift to renewable energy sources but also the safe decommissioning of older energy infrastructures and management of their toxic legacies. While the global scale of the decommissioning challenge is yet to be accurately quantified, the climate impacts are significant: each year, more than an estimated 29 million abandoned oil and gas wells around the world emit 2.5 million tons of methane, a potent greenhouse gas. In the US alone, at least 14 million people live within a mile of an abandoned oil or gas well, creating pollution that is concentrated among low-income areas and communities of colour. The costs involved in decommissioning projects are significant, raising urgent questions about responsibility and whether companies who have profited from the sale of extracted resources will be held liable for clean-up, remediation and management costs. Recognising these political goals and policy challenges, this article invites further research, scrutiny and debate on what would constitute the successful and safe decommissioning of sites affected by fossil fuel operations – with a particular focus on accountability, environmental inequality, the temporality of energy transitions, and strategies for phasing out or phasing down fossil fuel extraction.

Numerical simulation and experiments study on laser ablation decontamination for strontium and cesium from contaminated 316L stainless steels in spent nuclear fuel reprocessing

Radionuclides, 90Sr and 137Cs in this study, generated during spent fuel reprocessing will contaminate the structural material (316 L stainless steels), affecting the safe decommissioning of nuclear facilities. Laser ablation is a promising surface decontamination technology. In this study, the optimal laser parameters (the laser power of 16 W, the scanning speed of 175 mm/s, and the pitch spacing of 10 μm), used for laser ablation decontamination experiments for contaminated 316 L stainless steel in NaOH and HNO3 containing Sr and Cs which simulated the environments of spent nuclear fuel storage pond and Plutonium Uranium Redox Extraction (PUREX) process, were determined by numerical simulation with COMSOL Multiphysics. Laser induced breakdown spectroscopy (LIBS) was used to evaluate the decontamination performance of laser ablation, and no contaminant elements (Sr and Cs) were detected on the steel surface after laser ablation. Moreover, physicochemical technologies (SEM-EDS, XRD, and XPS) were used to characterize the surface chemical states and morphologies of contaminated steel before and after laser ablation. It was found that on the surface of the contaminated steel after laser ablation, δ-ferrite phase was detected, and the passive film was re-formed.

Cement-Enhancing Mechanical Behavior of Tailings Behind Upstream Tailings Dam for Safe Decommissioning

Cement-Enhancing Mechanical Behavior of Tailings Behind Upstream Tailings Dam for Safe Decommissioning

Applicability of abrasive waterjet cutting to irradiated graphite decommissioning

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

Asbestos Stocks and Flows Legacy in Australia

Information about asbestos stocks and flows is paramount for effective legacy management, both for understanding potential asbestos exposure risks from the different product types remaining in the built environment and proactive resource planning for their safe decommissioning, removal and disposal. This paper provides an overview of the Australian Stocks and Flows Model for Asbestos, a national model that provides best estimates to examine asbestos legacy stocks remaining in the built environment and flows to waste, now and into the future in Australia. The model was updated in 2021 to reflect new information from literature and input from industry experts and includes a Monte Carlo analysis to better reflect the range in the value estimates, as well as allowing for input of data from asbestos removal programs. Australia’s total asbestos stocks peaked at approximately 11 million tonnes in the 1980s. Over 95% of stocks comprise asbestos cement products, such as wall sheeting and water pipes. Australia’s current remaining asbestos stocks in the built environment are estimated at 6.2 million tonnes, with just under half of total consumption estimated to have gone to landfill as waste. The model can continue to be used with updated information to help track how much of Australia’s hazardous asbestos legacy is remaining and by how much it is reducing. The model can also be used to test scenarios and implications for predicted development trends and waste infrastructure needs. It is a valuable resource to assist with sustainable planning across a range of government departments that are responsible for managing asbestos waste in Australia.

Effect of strontium and cesium on corrosion and contamination of 316L stainless steels during spent nuclear fuel reprocessing

The high-yield fission products of spent nuclear fuel (SNF), 90Sr and 137Cs, can deposit on the surface of facilities for SNF reprocessing plant, threatening to their safe decommissioning. To select and develop the efficient decontamination technology, the experiments for corrosion and contamination with strontium and cesium of 316 L stainless steels were performed in simulated environments of alkaline storage pond and HNO3-based PUREX process. This study analyzed the mechanism and relations of corrosion and contamination with strontium and cesium on the surface of 316 L stainless steels by combining electrochemistry and ICP-MS for the first time. When steels were exposed to the environments of alkaline and low concentration of HNO3, the presence of Sr2+ and Cs + decreased these three properties (corrosion potential, holes formed by local severe dissolution and the stability of the passive film). The results of ICP-MS demonstrated that the contamination evolution of Sr and Cs had a similar trend, and that there were significantly different contamination behaviors of steels exposed to alkaline and acidic environments. Strontium and cesium mainly deposited in the passive film of steels in both nitric acid and alkaline environments. The contamination of steels in an alkaline environment was more serious than in all nitric acid environments.

Corrosion and Contamination of 316L Stainless Steel in Simulated HNO3-Based Spent Nuclear Fuel Reprocessing Environments with Cesium and Strontium

Austenitic 316L stainless steel is usually used as the structural material in spent fuel reprocessing facilities. However, the contamination of fission products from spent fuel (especially Cs and Sr) on the surface of facilities will greatly affect its safe decommissioning. In this study, for the first time, with the combination of characterizations of electrochemistry and laser-induced breakdown spectroscopy with two channels, the corrosion and contamination of 316L stainless steel were analyzed in simulated reprocessing environments with HNO3 concentrations of 3, 6, 8, and 12 M HNO3 containing Cs and Sr for 7, 14, 21, 30, and 60 days, respectively. The stability order of the passive film of 316L stainless steel in various environments was 12 M HNO3 > 8 M HNO3 > 3 M HNO3 > 6 M HNO3. Combined with other characterization methods, a mechanism was proposed. It was found that SrCO3, in the matrix, was the main contaminant in the initial 7 days in all environments due to a large number of holes. Then, the dissolution of contaminated steel surface led to the decrease of contaminant, and recontamination occurred after the repassivation of steel. These findings have significant implications for the development and selection of decontamination technology for the contaminated facilities made of 316L stainless steel.

Radiolysis Effects, Which should be Taken into Account for Safe Decommissioning of Fukushima-Daiichi Nuclear Power Station

Radiolysis Effects, Which should be Taken into Account for Safe Decommissioning of Fukushima-Daiichi Nuclear Power Station

Characterization of two types of cesium-bearing microparticles emitted from the Fukushima accident via multiple synchrotron radiation analyses

A part of radiocesium emitted during the Fukushima nuclear accident was incorporated in glassy water-resistant microparticles, called Type-A particles, which are spherical with ~ 0.1 to 10 µm diameter and ~ 10–2 to 102 Bq cesium-137 (137Cs) radioactivity; they were emitted from Unit 2 or 3 of the Fukushima Daiichi Nuclear Power Plant. Meanwhile, Type-B particles, having various shapes, 50–400 µm diameter, and 101–104 Bq 137Cs radioactivity, were emitted from Unit 1. The chemical properties of these radioactive particles have been reported in detail, but previous studies investigated only a small number of particles, especially Type-B particles. We tried to understand radioactive particles systematically by analyzing a large number of particles. Micro-X-ray computed tomography combined with X-ray fluorescence analysis revealed the presence of many voids and iron-rich part within Type-B particles. The 137Cs concentration (Bq mm–3) of Type-A particles was ~ 10,000 times higher than that of Type-B particles. Among the Type-B particles, the spherical ones had higher concentration of volatile elements than the non-spherical ones. These differences suggested that Type-A particles were formed through gas condensation, whereas Type-B particles were formed through melt solidification. These findings might contribute to the safe decommissioning of reactors and environmental impact assessment.

RADIATION HYGIENE AND SAFETY OF NUCLEAR INDUSTRY

This article covers basic issues of the radiation protection in nuclear industry. It contains an overview of history of the national nuclear industry including the creation of industry-specific facilities (research centers, medical units etc.). Main stages of the creating the regulatory system for radiation protection, starting from the beginning of the industrial radiation protection, stages of introducing exposure limits and implementation of the radiation protection system in international documents are described. In 1996, for the first time, radiation protection requirements in Russia were documented in the form of the Federal Law 3-FZ of 09.01.1996 “Radiation Protection of the Public". A new stage of updating the global methodological foundation of radiation protection began in 2007. IRCP recommendations of 2007 moved from the legacy practice and intervention approach focused on the process to the approach based on characteristics of exposure situation. The evolvement of new technologies (specifically, in the field of reactor engineering and used nuclear fuel) in recent years requires a special focus on the safety of the personnel and the public. This stipulates the necessity of the appropriate radiation protection support of activities for the safe implementation of modern technologies. Handling of spent nuclear fuel and generated radioactive wastes, safe decommissioning of radiation hazardous facilities, radiation protection during operation of radiation facilities in nonstandard conditions are all the issues requiring specific examination. Regulatory and procedural documents on radiation protection of the personnel and the public during development and implementation of new technologies have been developed and approved as a result of long-term work of scientists and other professionals.

Proposal for effective disposal options of very low level decommissioning waste

The safe decommissioning of nuclear power plants is one of the key issues in Korea due to the aging of NPPs. In this paper effective disposal options to dispose very low level decommissioning waste have been described through analysis of the silo for radioactive waste in several countries including Korea, specifically focused on disposal of the large component waste and other very low level waste generated from NPP decommissioning. As a result of the analysis, compact silo farm is proposed with two options: those are (1) ϕ22 × 23 m silo and (2) ϕ8.6 × 22 m silo. Each option has two disposal methods; one-piece disposal and disposal after volume reduction are considered as disposal method. Furthermore, arrangement of large component and the shape of silo are also proposed in this study. Only steam generator is considered to be disposed of as a large component because it can be categorized as very low level waste. In terms of disposal arrangement of steam generator, there are two ways: one-by-one disposal and disposal in pairs. To decrease the size of silo, both cylindrical and rectangular shapes are considered. In addition, the required number of silos and capacity are also estimated so that a comparison can be made between two options.

The Use of Simultaneous Thermal Analysis to Assess the Stability of Landfilled Municipal Solid Waste

There were presented results of study assessing the stability of the waste placed in the landfill-sites of municipal solid waste of the Perm Kray using thermal methods of analysis in air and argon. It was established that use of simultaneous thermal analysis method for samples of solid municipal waste, taken from landfills and dumps of solid municipal waste that are at different stages of the life cycle, allowing comprehensive determine the stability of place of landfilling. It was conducted testing methodology for assessing the degree of waste stability, based on the determination of the ratio of the specific enthalpy of dry matter waste (DM) and organic dry substance (ODS). It was justified by the possibility of using techniques to determine the degree of biochemical stability of waste at different stages of the landfill’s life cycle. It was established that at operated landfill-sites and dumps of solid waste DM and ODS were 0.45 or more, at reclaimed landfill-sites – 0.1 – 0.25, and at closed landfill-sites – less than 0.1. Developed a way to assess the stability of MSW samples allows to predict the potential emission of landfilled waste, which in turn can serve as a basis for the development of measures to reduce the negative impact of landfills on the environment, Technical and technological solutions for the remediation and safe decommissioning of solid waste dumping sites.

국내 원자로시설 해체계획서 세부 작성지침(안) 개발

원자로시설 해체계획서 작성은 원자로시설의 안전한 해체, 해체폐기물의 발생 최소화, 사람과 환경 보호 등을 위해 원자로시설 해체에 앞서 반드시 필요하다. 우리나라는 고리 1호기, 월성 1호기가 머지않은 시기에 해체를 맞이하게 되지만, 해체계획서를 작성하는데 필요한 관련 세부 지침이 미비한 상황이다. 이에 따라 본 논문에서는 IAEA의 원자력 국제 안전기준과 원자로시설 해체경험과 선진화된 규제체계를 보유하고 있는 미국, 영국, 프랑스의 해체계획서 작성 지침을 비교분석하고 국내 상황을 감안하여 원자로시설 해체계획서 세부 작성지침(안)을 개발하였다. 본 논문에서 개발한 작성지침(안)은 국내 원자로시설 해체 관련 법령을 정비하고 해체사업자가 원자로시설을 해체하고자 할 때 미리 준비사항을 알려주는 역할을 할 것으로 기대되고 있다. It is essential to prepare the decommissioning plan for a nuclear power plant (NPP) for the safe decommissioning of the NPP, minimization of the generation of decommissioning wastes, and protection of human beings and environment. Although Kori unit 1 and Wolsong unit 1 will be destined to their decommissioning in Korea in the near future. there is no provisons about preparing the decommissioning plan. In this paper, therefore, the draft guidelines of the decommissioning plan for a NPP were developed by considering the domestic situation, based on the comparative analyses of the regulatory guidelines of the decommissioning plan in U.S., U.K. and France. The draft guidelines are expected to play an important role to modify the domestic laws and regulations on the decommissioning of the NPP, and to give a license holder in charge of decommissioning the detailed instructions for preparing it in advance.

Toward Safe Decommissioning of Nuclear Facilities

Toward Safe Decommissioning of Nuclear Facilities

Development of standard evaluation plan for survey and investigation of residual radioactivity on site

The development of decommissioning technologies, including those for evaluation of the radioactivity inventory and the radiation dose, and a site investigation plan are needed to ensure safe decommissioning. The residual radioactivity should be measured and analysed to release a site for unrestricted use. The methods used for measurement of residual radioactivity have an effect on the workers' manpower and the decommissioning cost. So the development of the measurement methods of residual radioactivity with guidelines for a radiation survey and site investigation is needed to prepare for the future decommissioning of commercial Nuclear Power plants. The major considerations and characteristics of the parameters related to the decision of measurements points have been reviewed. The methods for controlling uncertainty and techniques to determine whether the measurement results achieve the survey objectives have been established.

ICONE19-43160 Development of Spent Ion Exchange Resin Processing in Nuclear Power Stations

Commercial operation of the Tsuruga nuclear power station Unit 1, owned by the Japan Atomic Power Company (JAPC), will be terminated in 2016. For safe decommissioning of the station, technologies for processing stored radioactive wastes such as spent Ion EXchange resin (IEX) and Filter Sludge (FS) have been jointly developed by JAPC and JGC Corporation. The Wet-Oxidation process (WetOx) was applied for decomposition of the spent IEX and FS, and Super Cement (SC) solidification was chosen for immobilization of the WetOx residue. Pilot scale tests for the WetOx treatment have been successfully conducted with simulated wastes. The WetOx residue is a concentrate with Suspended Solid (SS) and sodium sulfate, and was solidified by SC in lab scale cementation apparatus. The compressive strength of the solidified waste was confirmed to meet the desired level for radioactive waste disposal.