MANAGEMENT OF RADIOACTIVE WASTES IN CHINA

The policy and principles on management of radioactive wastes are stipulated.Cement solidification and bituminization unit has come into trial run. Solid radioactive waste is stored in tentative storage vault built in each of nuclear faCilities. Seventeen storages associated with applications of nuclear technology and radioisotopes have been built for provinces. Disposal of low and intermediate level radioactive wastes pursues the policy of 'regional disposal'. Four repositories have been planned to be built in northwest, southwest, south and east China respectively. A program for treatment and disposal of high level radioactive waste has been made.

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.

A novel method for gamma spectrum analysis of low-level and intermediate-level radioactive waste

The uncertainty of nuclide libraries in the analysis of the gamma spectra of low-and intermediate-level radioactive waste(LILW)using existing methods produces unstable results.To address this problem,a novel spectral analysis method is proposed in this study.In this method,overlapping peaks are located using a continuous wavelet transform.An improved quadratic convolution method is proposed to calculate the widths of the peaks and establish a fourth-order filter model to estimate the Compton edge baseline with the overlapping peaks.Combined with the adaptive sensitive nonlinear iterative peak,this method can effectively subtracts the background.Finally,a function describing the peak shape as a filter is used to deconvolve the energy spectrum to achieve accurate qualitative and quantitative analyses of the nuclide without the aid of a nuclide library.Gamma spectrum acquisition experiments for standard point sources of Cs-137 and Eu-152,a segmented gamma scanning experiment for a 200 L standard drum,and a Monte Carlo simulation experiment for triple overlapping peaks using the closest energy of three typical LILW nuclides(Sb-125,Sb-124,and Cs-134)are conducted.The results of the experiments indicate that(1)the novel method and gamma vision(GV)with an accurate nuclide library have the same spectral analysis capability,and the peak area calculation error is less than 4%;(2)compared with the GV,the analysis results of the novel method are more stable;(3)the novel method can be applied to the activity measurement of LILW,and the error of the activity reconstruction at the equivalent radius is 2.4%;and(4)The proposed novel method can quantitatively analyze all nuclides in LILW without a nuclide library.This novel method can improve the accuracy and precision of LILW measurements,provide key technical support for the reasonable disposal of LILW,and ensure the safety of humans and the environment.

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.

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.

Clays in radioactive waste disposal

Clays and argillites are considered in some countries as possible host rocks for nuclear waste disposal at great depth.The use of compacted swelling clays as engineered barriers is also considered within the framework of the multi-barrier concept.In relation to these concepts,various research programs have been conducted to assess the thermo-hydro-mechanical properties of radioactive waste disposal at great depth.After introducing the concepts of waste isolation developed in Belgium,France and Switzerland,the paper describes the retention and transfer properties of engineered barriers made up of compacted swelling clays in relation to microstructure features.Some features of the thermo-mechanical behaviors of three possible geological barriers,namely Boom clay（Belgium）,Callovo-Oxfordian clay（France） and Opalinus clay（Switzerland）,are then described,including the retention and transfer properties,volume change behavior,shear strength and thermal aspects.

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.

Implications of safety requirements for the treatment of THMC processes in geological disposal systems for radioactive waste

The mission of nuclear safety authorities in national radioactive waste disposal programmes is to ensure that people and the environment are protected against the hazards of ionising radiations emitted by the waste.It implies the establishment of safety requirements and the oversight of the activities of the waste management organisation in charge of implementing the programme.In Belgium,the safety requirements for geological disposal rest on the following principles:defence-in-depth,demonstrability and the radiation protection principles elaborated by the International Commission on Radiological Protection(ICRP).Applying these principles requires notably an appropriate identification and characterisation of the processes upon which the safety functions fulfilled by the disposal system rely and of the processes that may affect the system performance.Therefore,research and development(R&D)on safety-relevant thermo-hydro-mechanical-chemical(THMC)issues is important to build confidence in the safety assessment.This paper points out the key THMC processes that might influence radionuclide transport in a disposal system and its surrounding environment,considering the dynamic nature of these processes.Their nature and significance are expected to change according to prevailing internal and external conditions,which evolve from the repository construction phase to the whole heatingecooling cycle of decaying waste after closure.As these processes have a potential impact on safety,it is essential to identify and to understand them properly when developing a disposal concept to ensure compliance with relevant safety requirements.In particular,the investigation of THMC processes is needed to manage uncertainties.This includes the identification and characterisation of uncertainties as well as for the understanding of their safety-relevance.R&D may also be necessary to reduce uncertainties of which the magnitude does not allow demonstrating the safety of the disposal system.

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.

NEW IDEAS AND METHODS ON GEOMECHANICAL SAFETY ANALYSIS OF TOXIC AND RADIOACTIVE WASTE REPOSITORIES IN SALT MINE

In view of the need for geomechanical safety analysis of repositories in salt rock, failure criteria,creep rupture criteria,material models,pillar design methods and criteria for the assessment of barrier efficiency as well as investigations of the interaction between hydraulics and mechanics for the case of uncontrolled flooded repositories are necessary. The introduction of damage mechanics and of the Hou/Lux material model including damages into geomechanical safety analysis of repositories in salt rock can reduce some previous deficits in knowledge and modelling. This article will be as a part of geotechnical assessment to introduce the Hou/Lux material model,a new concept of hydro-mechanical coupling and a pillar design method as well as criteria for the assessment of efficiency of geological barriers.

Combustion synthesis of radioactive waste immobilization

Using chromium oxide (CrO3) as an oxidant, the immobilization of simulating radioactive waste in perovskite (CaTiO3) structure by a combustion synthesis (CS) method was tested. The products were characterized by Archimedes liquid displacement technique, microhardness technique, X-ray diffraction, and scanning electron microscopy. The leaching rate was measured by the method of MCC-1 or MCC-2. The primary results show that the CS method can be used to solidify the immobilizate waste effectively.

On the thermal impact on the excavation damaged zone around deep radioactive waste disposal

Clays and claystones are considered in some countries （including Belgium, France and Switzerland） as a potential host rock for high activity long lived radioactive waste disposal at great depth. One of the aspects to deal with when assessing the performance of the disposal is related to the effects of the temperature elevation due to the placement of exothermic wastes on the host rock. The potential effects of the thermal impact on the excavation damaged zone （EDZ） in the near field are another important issue that was the goal of the TIMODAZ European research project. In this paper, some principles of waste disposal in clayey host rocks at great depth are first presented and a series of experimental investigations carried out on specific equipment specially developed to face the problem are presented. Both drained and undrained tests have been carried out to investigate the drained thermal volume changes of clays and claystone and the thermal pressurization occurring around the galleries. The importance of proper initial saturation （under in situ stresses） and of satisfactory drainage conditions （in spite of the significantly low perme- ability of claystones） is emphasized, leading to the development of a new hollow cylinder apparatus. It is observed that claystones cannot be considered as overconsolidated clays given that they can exhibit, as the Callovo-Oxfordian claystone does, a thermoplastic contraction. Mechanical and thermal hardenings are however observed, extending the knowledge already gained on clays to claystones. A new method of determining the thermal pressurization coefficient in the laboratory is described and the data obtained allow completing existing data in the field. Finally, the hollow cylinder apparatus makes it possible to demonstrate that the good self-sealing properties of clays and claystones can be extended to temperature effects, an important conclusion in terms of performance assessment.

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.

Medium-Deep or Very Deep Disposal of Highly Radioactive Waste？

Commonly proposed concepts, like KBS-3V, for disposal of highly radioactive waste imply construction at medium depth （400-600 m） in granitic rock, which is excellent for constructing a stable repository. VDH （very deep boreholes） represent an altemative concept with the advantage that the rock is much less permeable and that the very salt, heavy groundwater is stagnant. Both require engineered barriers in the form of canisters and waste-embedding clay but for somewhat different purposes. Canisters are the most important waste-isolating barriers for KBS-3V but are less important for VDH. The waste-embedding clay is needed for preserving the KBS-3V canisters by being tight and ductile, but plays a minor role for the VDH. The backfilled deposition tunnels in a KBS-3V repository provide very limited hindrance of radionuclides to move to the biosphere while the clay seals of VDH effectively prevent possibly released radionuclides to reach up to the biosphere. Comparison of the KBS-3V and VDH concepts indicates that the last mentioned one has several advantages but that certain issues remain to be worked on for becoming a number one candidate.

Use of Low Radioactive Rare-earth Waste for Sewage Treatment

Low radioactive rare earth waste (containing 232 Th,specific activity 5 000 8 000 Bq·kg -1 ) were diluted 20 times by cement,sand and carbide ash and were made into special cement.The radioactivity of this special cement complied with the healthy protect standard for radioactive materials (GB6566 86).Test results showed that this special cement could lower COD,the degradation rate increased as the time went on.In acidic medium,this special cement could remove E Coli effectively.Applying aeration and adding lumps of cement,the degradation of COD versus time complied with Logistic model through fitting by computer.The two “S” curves indicated that aeration and adding lumps of cement had synergistic action on sewage treatment.

Indonesia’s Local Material Effect in Clay-Based Ceramic Filter Fabrication as an Alternative for Liquid Radioactive Waste Processing Material

One of the procedures to handle liquid radioactive waste is by filtration process. To do this process, suitable filter should be used because of radioactive nature of the waste. Ceramic filter is one of the suitable filters that could be used for this purpose. This paper will discuss about producing ceramic filter from local clay and test its performance. Performance of the filter is given by its flux, compressive strength, Decontamination Factor (DF) and adsorption efficiency. The results show that there are almost no effects of casting pressure on both flux and compressive strength of ceramic filter, but zeolite addition produces different effect. The higher concentration of zeolite will decrease the filter flux and increase filter compressive strength. The optimal composition from this research is 70% w/o clay-25% w/o zeolite-5% w/o charcoal. It has adsorption efficiency (60.36) and Decontamination Factor (2.52). Besides, Sr concentration after filtration is still higher than environmental standard for Sr-90 and more studies are still needed.

Selective recovery of Th(Ⅳ) from radioactive rare earth waste residue by utilizing MoS_(2)-modified ion-absorbed type rare earth tailings

In the process of extracting ion-absorbed rare earth ore(IREO),the production of radioactive waste is a major environmental concern.To address this issue,MoS_(2) was used to modify ion-absorbed rare earth tailings(RET) to synthesize a novel MoS_(2)@RET composite material for the effective handling of radioactive waste generated in IREO separation industry.The composite material was thoroughly characterized using various analytical techniques,including X-ray diffraction(XRD),X-ray photoelectron spectroscopy(XPS),thermogravimetry(TG),Fourier-transform infrared(FTIR),scanning electron microscopy(SEM),Brunaue r-Emmett-Teller(BET) and energy dispersive spectroscopy(EDS).By optimizing the operating parameters,the optimal experimental conditions were determined to be pH=3,contact time=60 min,liquid-solid ratio=6 g/L,and initial concentration=150 mg/L.The adsorption data fitted well with the pseudo second-order rate model.The thermodynamic parameters concerning the adsorption of Th(Ⅳ) were analyzed and computed.Langmuir isotherm model is a more fitting choice for the adsorption process compared to the Freundlich isotherm model.MoS_(2)@RET was used in the acid leachate of IREO waste residue,achieving the separation of Th and rare earth successfully.The mechanism of Th(Ⅳ) adsorption by MoS_(2)@RET was investigated,revealing that the adsorption process involves electrostatic interactions,chemical bonding,and redox reactions.The above research results indicate that MoS_(2)@RET composite materials have application potential in the sustainable treatment of IREO radioactive waste.