Effect of temperature on copper corrosion in high-level nuclear waste environment

The effect of temperature on the corrosion behavior of copper in simulated high-level nuclear waste environment wassystematically studied.Electrochemical methods,including electrochemical impendence spectra,Mott–Schottky technology,cyclicpolarization,and potentiostatic polarization,were employed to characterize the corrosion behavior of copper at different temperatures.Stereoscopic microscopy and scanning electron microscopy were used to examine the surface morphology,and X-ray photoelectronspectroscopy analysis was used to identify the composition of the passive film.The experimental results show that corrosionresistance of the passive film does not blindly decrease with the increase of temperature but increases at60°C owing to a compactouter layer;there is a potential for pitting corrosion,which decreases as the temperature increases.The main product of copper in ananaerobic aqueous sulfide solution is Cu2S but the content of CuS increases at higher temperatures.The whole passivation rangeshows p-type semiconductor characteristics and the magnitude of the acceptor density is1023cm-3,which increases with increasingtemperature.

Atomistic Simulation Study of Defect Structure of Zircon as a High-Level Nuclear Waste Host Form

A set of potential parameters for modeling zircon was obtained by atomistic simulation techniques and a reasonable structural model of zircon was established by fitting some important properties of zircon.Based on the equilibrium configuration of zircon, authors calculated the formation energies of basic point defects and intrinsic disorders. The heats of solution of substituting Pu for Zr showed that there was an immiscible gap at the composition of (Pu75%-Zr25%, in mole fraction), which suggests that the amount of Pu substituting for Zr in zircon be≤50%.

High-Level Nuclear Wastes and the Environment: Analyses of Challenges and Engineering Strategies

The main objective of this paper is to analyze the current status of high-level nuclear waste disposal along with presentation of practical perspectives about the environmental issues involved. Present disposal designs and concepts are analyzed on a scientific basis and modifications to existing designs are proposed from the perspective of environmental safety. A new concept of a chemical heat sink is introduced for the removal of heat emitted due to radioactive decay in the spent nuclear fuel or high-level radioactive waste, and thermal spikes produced by radiation in containment materials. Mainly, UO2 and metallic U are used as fuels in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements which would remain radioactive for 104 to 108years. Essential concepts and engineering strategies for spent nuclear fuel disposal are described. Conceptual designs are described and discussed considering the long-term radiation and thermal activity of spent nuclear fuel. Notions of physical and chemical barriers to contain nuclear waste are highlighted. A timeframe for nuclear waste disposal is proposed and time-line nuclear waste disposal plan or policy is described and discussed.

Active source seismic imaging on near-surface granite body:case study of siting a geological disposal repository for high-level radioactive nuclear waste

In order to research whether it is suitable to set a geological disposal repository for high-level radioactive nuclear waste into one target granite body,two active source seismic profles were arranged near a small town named Tamusu,Western China.The study area is with complex surface conditions,thus the seismic exploration encountered a variettraveltimey of technical difculties such as crossing obstacles,de-noising harmful scattered waves,and building complex near-surface velocity models.In order to address those problems,techniques including cross-obstacle seismic geometry design,angle-domain harmful scattered noise removal,and an acoustic wave equation-based inversion method jointly utilizing both the and waveform of frst arrival waves were adopted.The fnal seismic images clearly exhibit the target rock’s unconformable contact boundary and its top interface beneath the sedimentary and weathered layers.On this basis,it could be confrmed that the target rock is not thin or has been transported by geological process from somewhere else,but a native and massive rock.There are a few small size fractures whose space distribution could be revealed by seismic images within the rock.The fractures should be kept away.Based on current research,it could be considered that active source seismic exploration is demanded during the sitting process of the geological disposal repository for nuclear waste.The seismic acquisition and processing techniques proposed in the present paper would ofer a good reference value for similar researches in the future.

High-level radioactive waste disposal in China: update 2010

For geological disposal of high-level radioactive waste （HLW）, the Chinese policy is that the spent nuclear fuel （SNF） should be reprocessed first, followed by vitrification and final disposal. The preliminary repository concept is a shaft-tunnel model, located in saturated zones in granite, while the final waste form for disposal is vitrified high-level radioactive waste. In 2006, the government published a long-term research and development （R＆D） plan for geological disposal of high-level radioactive waste. The program consists of three steps： （1） laboratory studies and site selection for a HLW repository （2006-2020）; （2） underground in-situ tests （2021-2040）; and （3） repository construction （2041-2050） followed by operation. With the support of China Atomic Energy Authority, comprehensive studies are underway and some progresses are made. The site characterization, including deep borehole drilling, has been performed at the most potential Beishan site in Gansu Province, Northwestern China. The data from geological and hydrogeological investigations, in-situ stress and permeability measurements of rock mass are presented in this paper. Engineered barrier studies are concentrated on the Gaomiaozi bentonite. A mock-up facility, which is used to study the thermo-hydro-mechano-chemical （THMC） properties of the bentonite, is under construction. Several projects on mechanical properties of Beishan granite are also underway. The key scientific challenges faced with HLW disposal are also discussed.

The Beishan underground research laboratory for geological disposal of high-level radioactive waste in China:Planning, site selection,site characterization and in situ tests

With the rapid development of nuclear power in China, the disposal of high-level radioactive waste（HLW） has become an important issue for nuclear safety and environmental protection. Deep geological disposal is internationally accepted as a feasible and safe way to dispose of HLW, and underground research laboratories（URLs） play an important and multi-faceted role in the development of HLW repositories. This paper introduces the overall planning and the latest progress for China＇s URL. On the basis of the proposed strategy to build an area-specific URL in combination with a comprehensive evaluation of the site selection results obtained during the last 33 years, the Xinchang site in the Beishan area,located in Gansu Province of northwestern China, has been selected as the final site for China＇s first URL built in granite. In the process of characterizing the Xinchang URL site, a series of investigations,including borehole drilling,geological mapping, geophysical surveying,hydraulic testing and in situ stress measurements, has been conducted. The investigation results indicate that the geological,hydrogeological, engineering geological and geochemical conditions of the Xinchang site are very suitable for URL construction. Meanwhile, to validate and develop construction technologies for the Beishan URL, the Beishan exploration tunnel（BET）, which is a 50-m-deep facility in the Jiujing sub-area, has been constructed and several in situ tests, such as drill-and-blast tests, characterization of the excavation damaged zone（EDZ）, and long-term deformation monitoring of surrounding rocks, have been performed in the BET. The methodologies and technologies established in the BET will serve for URL construction.According to the achievements of the characterization of the URL site, a preliminary design of the URL with a maximum depth of 560 m is proposed and necessary in situ tests in the URL are planned.

Selection and Basic Properties of the Buffer Material for High-Level Radioactive Waste Repository in China

Radioactive wastes arising from a wide range of human activities are in many different physical and chemical forms, contaminated with varying radioactivity. Their common features are the potential hazard associated with their radioactivity and the need to manage them in such a way as to protect the human environment. The geological disposal is regarded as the most reasonable and effective way to safely disposing high-level radioactive wastes in the world. The conceptual model of geological disposal in China is based on a multi-barrier system that combines an isolating geological environment with an engineered barrier system. The buffer is one of the main engineered barriers for HLW repository. It is expected to maintain its low water permeability, self-sealing property, radio nuclides adsorption and retardation properties, thermal conductivity, chemical buffering property, canister supporting property, and stress buffering property over a long period of time. Bentonite is selected as the main content of buffer material that can satisfy the above requirements. The Gaomiaozi deposit is selected as the candidate supplier for China＇s buffer material of high level radioactive waste repository. This paper presents the geological features of the GMZ deposit and basic properties of the GMZ Na-bentonite. It is a super-large deposit with a high content of montmorillonite （about 75 %）, and GMZ-1, which is Na-bentonite produced from GMZ deposit is selected as the reference material for China＇s buffer material study.

Design and development of large-scale in-situ PRACLAY heater test and horizontal high-level radioactive waste disposal gallery seal test in Belgian HADES

In Belgium,the Boom clay was selected as a potential host formation for the disposal of high-level radioactive waste（HLW）.To demonstrate the suitability of Boom clay for bearing thermal load induced by the HLW,a large-scale in-situ heater test,called PRACLAY heater test,will be conducted in the underground research laboratory（URL） in Mol.Owing to the limitations of the test（a short period of time compared with that considered in a real repository,different boundary conditions,etc.）,the test is designed to simulate,in a conservative way,the most critical state and phenomena that could occur in the host rock.The PRACLAY gallery was excavated at the end of 2007;the heating phase will begin in 2010 and will last for at least 10 years.The PRACLAY gallery itself leaves an opportunity to study the possibilities of sealing a disposal drift in Boom clay and testing the feasibility of hydraulic cut-off of any preferential pathway to the main access gallery through the excavation damage zone（EDZ） and the lining with a seal in a horizontal drift（horizontal seal）.Indeed,this is a generic problem for all deep geological disposal facilities for HLW.An annular seal made of compacted swelling bentonite will be installed in the front of the heated part of the PRACLAY gallery for these purposes.This paper provides detailed considerations on the thermo-hydro-mechanical（THM） boundary conditions for the design of the PRACLAY heater test and the seal test with the support of numerical calculations.It is believed that these important items considered in the PRACLAY heater test design also constitute key issues for the repository design.The outcome of the PRACLAY heater test will be an important milestone for the Belgian repository design.

Main outcomes from in situ thermo-hydro-mechanical experiments programme to demonstrate feasibility of radioactive high-level waste disposal in the Callovo-Oxfordian claystone

In the context of radioactive waste disposal,an underground research laboratory(URL)is a facility in which experiments are conducted to demonstrate the feasibility of constructing and operating a radioactive waste disposal facility within a geological formation.The Meuse/Haute-Marne URL is a sitespecific facility planned to study the feasibility of a radioactive waste disposal in the Callovo-Oxfordian(COx)claystone.The thermo-hydro-mechanical(THM)behaviour of the host rock is significant for the design of the underground nuclear waste disposal facility and for its long-term safety.The French National Radioactive Waste Management Agency(Andra)has begun a research programme aiming to demonstrate the relevancy of the French high-level waste(HLW)concept.This paper presents the programme implemented from small-scale(small diameter)boreholes to full-scale demonstration experiments to study the THM effects of the thermal transient on the COx claystone and the strategy implemented in this new programme to demonstrate and optimise current disposal facility components for HLW.It shows that the French high-level waste concept is feasible and working in the COx claystone.It also exhibits that,as for other plastic clay or claystone,heating-induced pore pressure increases and that the THM behaviour is anisotropic.

Effect of drying cracks on swelling and self-healing of bentonite-sand blocks used as engineered barriers for radioactive waste disposal

Experiments were conducted to evaluate the healing of drying cracks in air-dried bentonite-sand blocks after hydration and swelling in groundwater,providing justifications to simplify the protection of blocks prior to installation in a high-level radioactive waste repository.Synthetic groundwater was prepared to represent the geochemistry of Beishan groundwater,and was used to hydrate the blocks during the swelling pressure and swelling strain measurements,as Beishan is the most promising site for China's repository.Healing of the surface cracks was recorded by photography,and healing of the internal cracks was visualized by CT images and hydraulic conductivity of air-dried blocks.The results indicate that the maximum swelling pressure and swelling strain are primarily affected by the geochemistry of Beishan groundwater,but not affected by the drying cracks.The maximum swelling pressure and swelling strain of air-dried blocks are comparable to or even higher than the pressure and strain of fresh blocks.The maximum swelling pressure measured in strong(i.e.high ion strength)Beishan groundwater was 44%of the pressure measured in deionized(DI)water,and the maximum swelling strain was reduced to 23%of the strain measured in DI water.Nevertheless,the remained swelling of the blocks hydrated in strong Beishan groundwater was sufficient to heal the surface and internal drying cracks,as demonstrated by the pictures of surface cracks and CT images.The hydraulic conductivity of the air-dried block permeated with strong groundwater was comparable(3.7×higher)to the hydraulic conductivity of the fresh block,indicating the self-healing of drying cracks after hydration and swelling in groundwater.A simplified method of protecting the block with plastic wraps before installation is recommended,since the remained swelling of the block hydrated in Beishan groundwater is sufficient to heal the drying cracks.

Experimental study on the physico-mechanical properties of Tamusu mudstone — A potential host rock for high-level radioactive waste in Inner Mongolia of China

Tamusu mudstone, located in Bayin Gobi Basin in Inner Mongolia of China, has been selected as a potential host rock for high-level radioactive waste(HLW) disposal in China. A series of tests has been carried out, including X-ray diffraction(XRD) tests, scanning electron microscopy(SEM) tests, disintegration tests, permeability tests and triaxial compression tests, to estimate the physico-mechanical properties of Tamusu mudstone in this work. The mineral composition of Tamusu mudstone was analyzed and it was considered as a stable rock due to its low disintegration rate, i.e. approximately 0.11%after several wet/dry cycles. Based on the results of permeability test, it was found that Tamusu mudstone has a low permeability, with the magnitude of about 10—20m^(2). The low permeability makes the mudstone well prevent nuclide migration and diffusion, and might be influenced by temperature.The triaxial tests show that Tamusu mudstone is a stiff mudstone with high compressive strength, which means that the excavation disturbed zone would be smaller compared to other types of mudstone due to construction and operation of HLW repositories. Finally, the properties of Tamusu mudstone were compared with those of Opalinus clay, Callovo-Oxfordian(COx) argillite, and Boom clay to further discuss the possibility of using Tamusu mudstone as a potential nuclear waste disposal medium.

On area-specific underground research laboratory for geological disposal of high-level radioactive waste in China

Underground research laboratories （URLs）, including ＂generic URLs＂ and ＂site-specific URLs＂, are un- derground facilities in which characterisation, testing, technology development, and/or demonstration activities are carried out in support of the development of geological repositories for high-level radioactive waste （HLW） disposal. In addition to the generic URL and site-specific URL, a concept of ＂areaspecific URL＂, or the third type of URL, is proposed in this paper. It is referred to as the facility that is built at a site within an area that is considered as a potential area for HLW repository or built at a place near the future repository site, and may be regarded as a precursor to the development of a repository at the site. It acts as a ＂generic URL＂, but also acts as a ＂site-specific URL＂ to some extent. Considering the current situation in China, the most suitable option is to build an ＂area-specific URL＂ in Beishan area, the first priority region for China＇s high-level waste repository. With this strategy, the goal to build China＇s URL by 2020 mav be achieved, but the time left is limited.

Corrosion behavior of high-level waste container materials Ti and Ti-Pd alloy under long-term gamma irradiation in Beishan groundwater

Titanium and titanium-palladium alloys are important potential materials for nuclear waste container,which will endure both intenseγ-irradiation and groundwater erosion.Therefore,it is very important to investigate the corrosion behavior of the container materials.In this research,the cumulative dose effect of TA8-1 type titanium-palladium alloy(TA8-1)and TA2-type pure titanium(TA2)underγ-irradiation was studied based on the geological disposal of nuclear wastes.The irradiation experiments were performed at room temperature using^(60)Co gamma sources with a 5.0-kGy·h^(-1)intensity for 40,80 or 160 days,respectively.The pH value and conductivity of Beishan groundwater were investigated.The results showed that the pH value changed from alkaline(8.22)to acidic(2.46 for TA8-1 and 2.44 for TA2),while the un-irradiated solution remained alkaline(8.17 for TA8-1 and 8.20 for TA2)after 160 days.With the increase of irradiation dose,the conductivity increases rapidly and then tends to become stable,which indicates that the titanium dioxide corrosion layer formed on the surface of the sample surface effectively prevents further corrosion.Meanwhile,XRD and SEM-EDS analysis results show that the main components of corrosion products are TiO_(2) and TiO.The titanium on the surface of the sample is oxidized,resulting in slight uneven local corrosion.The results show that TA8-1 and TA2 are suitable to be used as candidate materials for high-level waste(HLW)disposal containers due to their excellent performance under long-term and high-dose irradiation corrosion.

Self-propagating High-temperature Synthesis of Sm and Zr Co-doped Gd2Ti2O7 Pyrochlore Ceramics as Nuclear Waste Forms

We reported a rapid synthesis of Sm^(3+)/Zr^(4+)co-doped Gd2Ti2O7 pyrochlore simulated nuclear wastes solidification by self-propagation plus quick pressing technique.With increment excess contents of Sm2O3 and ZrO2 from 0 to 10wt%,the phase composition of the products is a mixed phase of pyrochlore structure and defective fluorite structure by X-ray diffraction(XRD)analysis and Raman spectrum.In addition,the SEM results demonstrate the fracture surface and microstructure of Gd2Ti2O7-based pyrochlore.The densified pyrochlore waste form exhibits high bulk density of 5.56 g·cm^(-3) and vickers hardness of 11.20±0.2 GPa.The leaching tests show that the elemental leaching rates of Gd,Sm,and Cu after 42 days are 1.92×10^(-4),1.51×10^(-4),and 3.90×10^(-3) g·m^(-2)·d^(-1),respectively.

Recent progress and perspective on batteries made from nuclear waste

Sustainable energy sources are an immediate need to cope with the imminent issue of climate change the world is facing today.In particular,the long-lasting miniatured power sources that can supply energy continually to power handheld gadgets,sensors,electronic devices,unmanned airborne vehicles in space and extreme mining are some of the examples where this is an acute need.It is known from basic physics that radioactive materials decay over few years and some nuclear materials have their half-life until thousands of years.The past five decades of research have been spent harnessing the decay energy of the radioactive materials to develop batteries that can last until the radioactive reaction continues.Thus,an emergent opportunity of industrial symbiosis to make use of nuclear waste by using radioactive waste as raw material to develop bat-teries with long shelf life presents a great opportunity for sustainable energy resource development.However,the current canon of research on this topic is scarce.This perspective draws fresh discussions on the topic while highlighting future directions in this wealthy arena of research.Graphical abstract A long-lasting miniaturised nuclear battery utilising 14C radioactive isotope as fuel.

Recent advances of annular centrifugal extractor for hot test of nuclear waste partitioning process

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Collimated LaBr_3 detector response function in radioactivity analysis of nuclear waste drums

The properties of a Lanthanum bromide(LaBr3)detector and its response functions were investigated via experiments and simulations in this paper.The LaBr3detector had good relative energy resolution and higher efficiency than a high-purity germanium detector.Monte Carlo and other numerical methods were used to calculate the efficiencies of a LaBr3detector with a square collimation window.A model of the numerical method was established based on a pure geometric model that was consistent with the experimental situation.The results showed that the detector response functions calculated by these methods were in great agreement with experimental results.

GALERKIN METHOD FOR COMPRESSIBLE FLOW OF CONTAMINATION FROM NUCLEAR WASTE WITH MOLECULAR DIFFUSION AND DISPERSION

Abstract A system of quasilinear coupled equations which arise from simulation of contamination of geologic nulear waste in porous media is studied. We’ll discuss Galerkin method for the model of compressible flow with molecular diffusion and dispersion. Some new techniques are introcued to error analysis. Only one dimensional case is considered. The optimal error estimate in both L^2 and H^1 is proved. A contribution of this paper is how the dispersion term can be handled,

Study on the residence time of deep groundwater for high-level radioactive waste geological disposal

Residence time of deep groundwater is one of the most important parameters in safety and performance assessment for high-level radioactive waste geological disposal. In this study, we collected the deep groundwater samples of Jijicao in Gansu Beishan pre-selected region. The deep groundwater residence time at two depths estimated by Helium-4 accumulation method were 3.8 ka and 5.0 ka respectively upon measurement and calculation, which indicates that the deep groundwater is not derived from the deep crust circulation process. Hence, deep groundwater is featured with long residence time as well as slow circulation and update rate, and such features are conductive to the safe disposal of high-level radioactive waste.

Numerical analysis of thermal process in the near field around vertical disposal of high-level radioactive waste

For deep geological disposal of high-level radioactive waste(HLW)in granite,the temperature on the HLW canisters is commonly designed to be lower than100fiC.This criterion dictates the dimension of the repository.Based on the concept of HLW disposal in vertical boreholes,thermal process in the nearfield(host rock and buffer)surrounding HLW canisters has been simulated by using different methods.The results are drawn as follows:(a)the initial heat power of HLW canisters is the most important and sensitive parameter for evolution of temperaturefield;(b)the thermal properties and variations of the host rock,the engineered buffer,and possible gaps between canister and buffer and host rock are the additional key factors governing the heat transformation;(c)the gaps width and thefilling by water or air determine the temperature offsets between them.