Swelling behaviors of GMZ bentonite–sand mixtures inundated in NaCl–Na2SO4 solutions

In this study, the swelling behaviors of compacted GMZ bentonite–sand mixtures inundated in NaCl–Na₂SO₄ solutions are investigated and the influence of chemical solutions on the swelling behaviors of GMZ bentonite–sand mixtures as backfill/buffer material in China for high level radioactive waste (HLW) is investigated. The sand addition ratios of the bentonite–sand mixtures are 0%, 20%, 30% and 50%, and the total dissolved solids (TDS) of the NaCl–Na₂SO₄ (NaCl:Na₂SO₄ = 2:1 by mass) solution are 0.5, 1.0, 3.0, 6.0 and 12.0 g/L (pH 7.1). The specimens of bentonite–sand mixtures for swelling tests are prepared by static-compaction to various dry densities, ranging from 1.50 to 1.90 g/cm³.Test results indicate that liquid limit (LL) and plasticity limit (PL), swell time, maximum swelling pressure and maximum swelling strain decrease with the increase of TDS for GMZ bentonite–sand mixtures. All of the LL, PI and maximum swelling strain are decreased exponentially with TDS increase: very quickly as TDS < 3.0 g/L, slowly as TDS = 3.0–6.0 g/L and almost stabilized as TDS > 6.0 g/L. The maximum swelling pressure shows a linear reduction with the TDS increasing, but the pure bentonite indicates a high sensitivity than the bentonite–sand mixtures with 30% sand addition ratio. As NaCl–Na₂SO₄ (TDS = 0.5 g/L) solution was used according to the ground water, with initial dry density of 1.70 g/cm³, the maximum swelling pressure of specimens decrease exponentially while the maximum strain decrease linearly with the increase of sand addition. With 30% sand addition in 0.5 g/L NaCl–Na₂SO₄ solution, the maximum swelling pressure increase exponentially while the maximum strain increase linearly with the increase of initial dry density.Compared with the pure bentonite, bentonite–sand mixtures show a better tolerance withstanding the chemical attack to ground water chemistry because of the replacement of some quantity of expansive clay by quartz sand in the mixtures.     

Progress of a cost optimization for an HLW repository in Korea

This paper provides the results of a cost optimization for a CANDU spent fuel canister as well as the operational duration of an HLW repository. From the design change of an advanced-CANDU spent fuel canister, the overall costs were expected to be reduced by 124 MEUR in the case of disposing of 36,000 tU in an HLW repository, and it was also found that the optimal operational duration for an HLW repository was 83 years, to minimize the total cost. But this operational duration was only calculated from the aspect of cost benefits with economics' perspectives.We confirmed that the canister and operational duration are the dominant cost drivers for surface facilities and underground facilities for a cost optimization, respectively. Especially, the manufacturing method of an outer canister using the cold spray coating technique which was developed through collaboration with a domestic company is suggested to minimize the overall costs.     

缓冲/回填材料——内蒙古高庙子膨润土性能研究

研究按缓冲回填材料的性能要求,对内蒙古高庙子膨润土的物质组成与结构、物理化学性质、水理性质等进行了系统研究;对添加剂改善膨润土的压实性能进行了探索性实验研究。结果表明,高庙子钠基膨润土以蒙脱石为主要矿物成分,具有良好的吸附性、膨胀性、阳离子交换性和热稳定性;石英砂添加剂可以有效地改善膨润土的压实性能;进一步论证了内蒙古高庙子膨润土矿床可以作为我国高放废物地质处置库缓冲/回填材料的供给基地。     

高放废物地质处置库缓冲／回填材料性能测定

根据矿床位置、交通、矿区地质特征、矿床成因、矿床储量、我自然地理与开采技术条件等因素综合对比研究，研究内蒙古兴和县高庙子膨润土矿床为我国高放废物地质处置库缓冲／回填材料供给基地的首选矿床。本文对主要矿层进行了物质成分、物理化学性能等方面的研究。对测试结果分析可知，高庙子膨润土的蒙脱石含量较高，物理化学性能和物理水理性质较好。因此，高庙子膨润土矿床作为高放废物地质处置库缓冲／回填材料的供给基地是可行     

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.     

Calculation of maximum release rates in alternative design changes in the thickness of the buffer for the engineered barrier system (EBS) in deep repository by using Amber code

The multibarrier concept forms the basis for geological disposal concepts in most countries and the guideline states that research and development should aim to demonstrate the feasibility of constructing an engineered barrier system (EBS) which is appropriate for the range of relevant geological conditions. The multibarrier system (EBS) and its functions consisting of the glass waste form, overpack and buffer material was located in a sufficiently stable geological environment. When an overpack comes into contact with groundwater it will start to corrode. The wall thickness will then gradually reduce, and the overpack will eventually fail mechanically when its structural strength can no longer support the stress imposed by the surrounding environment. The requirements that influence the thickness of buffer include nuclide migration retardation and heat conductivity, as well as stress buffering capability, self-sealing ability and workability. The migration retardation function is assumed to be the most important of all these requirements with respect to setting the appropriate thickness of buffer. A consideration of these effects and relationship between buffer thicknesses has determined that a reasonable thickness for the buffer is between 400 mm and 700 mm [AECL, H12]. Therefore, the design thickness of buffer material can range from 0.4 m to 0.7 m to account for manufacturing and stress buffering. In this alternative design case, the thickness of the buffer material is set to 0.4 m, 0.5 m, 0.6 m and 0.7 m. The nuclide migration properties of the buffer material are assumed to be the same (PNC, Development and Management of the Technical Knowledge Base for the Geological Disposal of HLW, Supporting Report 2: “Repository” Engineering Technology). The results of calculation are presented that some nuclides such as Se-79, Tc-99, Pd-107, Th-233, U-236, Pb-210, Ra-226 and Np-237 virtually unchanged in case the maximum release rate from EBS corresponding to change thickness of buffer material. Some nuclides such as Cs-135, Nb-94, Nb-93 m, Zr-93, Sn-126, Th-230, Ph-240, Pu-242, U-233, Ac-227, Pa-231 and Th-229 are very little greater for 40 cm, 50 cm and 60 cm in the maximum release rate compared with 70 cm. Maximum release of nuclides U-235, U-234 and U-238 increases in case of 50 cm and 60 cm thickness of buffer and in case 40 cm are the same as 70 cm thickness because the amount of their parents in case 40 cm will decrease before decay and in case 70 cm amount of these nuclides will decrease due to decayed to other nuclides before release from the buffer, then maximum decay happened in case 50 cm. The maximum release rates of short-lived nuclides such as Cm-245, Am-243, Cm-245, Am-241, Pu-241 and Pu-239 increase significantly due to less decay occurring during the reduced buffer transit time.     

Engineering-scale test on the thermal–hydro–mechanical behaviors in the clay barrier of a HLW repository

The thermo–hydro–mechanical (T–H–M) behaviors of a clay barrier are of importance from a performance and safety viewpoint of the engineered barrier system (EBS) for a high-level waste (HLW) repository. An engineering-scale test was carried out to investigate the T–H–M behaviors in the buffer of the Korean reference disposal system (KRS). The test started on May 31, 2005 and is still in operation. The experimental data obtained allowed a preliminary and qualitative interpretation of the T–H–M behavior in bentonite blocks. The temperature was higher as it became closer to the heater, while it became lower as it was farther away from the heater. The water content had a higher value in the part close to the hydration surface than that in the heater part. The relative humidity data suggested that a hydration of the bentonite blocks might occur by different drying–wetting processes, depending on their position. The total pressure was continuously increased by the evolution of the saturation front in the bentonite blocks and thereby the swelling pressure. There was also a contribution of the thermal expansion of the bentonite blocks near the heater and the capillary force in the dry bentonite blocks which the water did not reach from the hydration surface.     

Ca^(2+)盐溶液在高庙子膨润土中的扩散研究北大核心CSCD

高放废物地质处置库近场地下水可能会对处置库内的屏障体系产生影响,降低处置库的安全稳定。为研究地下水中盐离子在处置库内缓冲回填体系的扩散规律,本文开展了静态无外荷载条件下内蒙古高庙子(GMZ)膨润土在Ca^(2+)盐溶液中自发渗吸的吸附扩散室内试验。从土的微观结构和经典扩散理论对Ca^(2+)在不同干密度和初始饱和度的膨润土试样中的自发扩散规律进行了分析。研究结果表明,在膨润土初始饱和度相同的情况下,试样阻滞系数随其干密度增加而增大,此时Ca^(2+)的扩散能力减弱;当膨润土干密度相同时,随着初始饱和度的增加基质吸力作用减弱,阻滞系数减小,Ca^(2+)的扩散能力减弱。     

中国高放废物地质处置21世纪进展

21世纪近20年,我国高放废物深地质处置进入了一稳步发展的新阶段,在法律法规、技术标准、战略规划、选址和场址评价、工程屏障研究、处置库和地下实验室概念设计、核素迁移和安全评价研究等方面取得了显著进展。其主要亮点包括颁布了《中华人民共和国放射性污染防治法》和《中华人民共和国核安全法》,制定了《高放废物地质处置研究开发规划指南》,颁布了《高放废物地质处置设施选址》核安全导则,确定了2020年前开工建设地下实验室、2050年建成高放废物处置库的目标,甘肃北山预选区被确定为我国高放废物地质处置库首选预选区,建立了场址评价方法技术体系,确定了内蒙古高庙子膨润土为我国高放废物处置库的首选缓冲回填材料,建立了我国首台缓冲回填材料热 水-力-化学耦合条件下特性研究大型实验台架（China-Mock-Up）,获得了一批关键放射性核素的迁移行为数据,开展了初步的安全评价,完成了地下实验室安全技术研究。确定甘肃北山的新场为我国高放废物地质处置地下实验室的场址。2019年5月6日,国家国防科工局批复中国北山高放废物地质处置地下实验室工程建设立项建议书,标志着我国高放废物地质处置正式进入地下实验室阶段。这一系列工作进展和取得的成绩为我国2020年开工建设地下实验室、掌握高放废物地质处置技术奠定了坚实的基础。     

缓冲/回填材料--膨润土研究国际进展

缓冲材料是高放废物地质处置库多重屏障系统重要组成部分。本文从膨润土特性、气体渗透性、膨润土中有机物、微生物腐蚀、孔隙水化学、蒙脱石向伊利石转化、核素迁移等方面简要总结了该领域的一些研究进展，旨在推动我国在这一领域的研究走向深入。目前，国内的工作主要集中于材料物理性能的测试，作者期待国家有关部门能加大经费支持力度。以推动这一领域的研究进展。确保高放废物的安全处置，为能源工业发展保驾护航。     

Calcium carbonate precipitation in compacted bentonite using electromigration reaction method and its application to estimate the ion activity coefficient in the porewater

In the safety assessment of radioactive waste disposal, it is critical to understand the porewater chemistry in compacted bentonite in order to predict long-term migration behavior of radionuclides in the engineered barrier. This study estimates the activity coefficients of dissolved ions in the porewater of compacted bentonite from the concentrations of ions at which CaCO₃ precipitation occurred. Solutions containing CaCl₂ and NaHCO₃ were introduced under electrical potential gradient from the opposite sides of the compacted Na-bentonite packed at the dry density of 1.0kg/dm³. After the electromigration, the spatial distribution of ions along the compacted bentonite sample was determined. Sequential extraction method was developed to distinctly determine the concentrations of free ions in the porewater and in solid phase in bentonite. The results show that the exchangeable Na⁺ ions were progressively replaced by the incoming Ca²⁺ ions, and the compacted bentonite sample can be divided into three zones: Ca-, Ca-/Na-, and Na-bentonite zones. Precipitates of CaCO₃ were observed both in Ca- and Ca/Na-bentonite zones. The experimentally determined activity coefficients were at least smaller by a factor of 3 compared to the theoretical approximation calculated using PHREEQC code assuming dilute-solution conditions with no electrostatic interactions between ions and bentonite surface.     

Tritium release experiments on Li4SiO4 pebbles from TRINPC-I experiments: Effects of water adsorption and hydrogen gas

Out-of-pile tritium release experiments under different water uptake contents and purge gas chemistry were performed on Li₄SiO₄. Water measurement was performed on samples under different experimental procedures. It was found that water was adsorbed on the sample during its transferring and storage process. A strong dependence of tritium release behavior on water uptake was determined. By doping H₂ in the sweep gas, the formation of water in orthosilicate was observed in addition to the isotope exchange reaction with H₂ gas. Thermal desorption peaks of the water formation reaction and H₂ isotope exchange reaction appeared at 668 °C and 463 °C, respectively, at ramping rate of 5 °C/min.     

Hydrogen incorporation in tungsten deposits growing by deuterium plasma sputtering

Tungsten deposits were produced by sputtering method using hydrogen isotope RF plasma, and the density and the incorporated components in the deposits were investigated. The density changed in the range from 14.2 g/cm³ to 6.1 g/cm³, and hydrogen isotope retention changed in the range from 0.25 to 0.05 as (H + D)/W by the difference of deposition conditions. Both the density and hydrogen isotope retention tended to decrease with an increase of pressure. Even though a deuterium gas was used for producing tungsten deposits, not only deuterium but also hydrogen, oxygen and water vapor were incorporated in the deposits. It is considered that the incorporation of these components originated in water vapor unintentionally existing in the vacuum chamber.     

Osmosis-induced water uptake by Eurobitum bituminized radioactive waste and pressure development in constant volume conditions

The chemo-hydro-mechanical (CHM) interaction between swelling Eurobitum radioactive bituminized waste (BW) and Boom Clay is investigated to assess the feasibility of geological disposal for the long-term management of this waste. These so-called compatibility studies include laboratory water uptake tests at the Belgian Nuclear Research Center SCK?CEN, and the development of a coupled CHM formulation for Eurobitum by the International Center for Numerical Methods and Engineering (CIMNE, Polytechnical University of Cataluña, Spain).In the water uptake tests, the osmosis-induced swelling, pressure increase and NaNO₃ leaching of small cylindrical BW samples (diameter 38 mm, height 10 mm) is studied under constant total stress conditions and nearly constant volume conditions; the actual geological disposal conditions should be intermediate between these extremes. Two nearly constant volume tests were stopped after 1036 and 1555 days to characterize the morphology of the hydrated BW samples and to visualize the hydrated part with microfocus X-ray Computer Tomography (μCT) and Environmental Scanning Electron Microscopy (ESEM). In parallel, a coupled CHM formulation is developed that describes chemically and hydraulically coupled flow processes in porous materials with salt crystals, and that incorporates a porosity dependent membrane efficiency, permeability and diffusivity.When Eurobitum BW is hydrated in (nearly) constant volume conditions, the osmosis-induced water uptake results in an increasing pressure to values that can be (in theory) as high as 42.8 MPa, being the osmotic pressure of a saturated NaNO₃ solution. After about four years of hydration in nearly constant volume water uptake tests, pressures up to 20 MPa are measured. During this hydration period only the outer layers with a thickness of 1–2 mm were hydrated (as derived from μCT and ESEM analyses), and only about 10–20% of the initial NaNO₃ content was released by the samples. In the studied test conditions, the rates of water uptake and NaNO₃ leaching are low because of the low porosity, and thus low permeability, of the hydrated BW samples in combination with a highly efficient semi-permeable bitumen membrane. In contrast to the hydration in free swelling conditions, the increase in porosity is limited by the high pressures in the nearly constant volume tests. Furthermore, at the interface with the stainless steel filters, a low permeable re-compressed bitumen layer is formed, as observed on the ESEM images.The experimental results of pressure increase and NaNO₃ leaching, as well as observations on μCT and ESEM images (e.g. compression of leached layers, high dissolved NaNO₃ concentration in hydrated BW after about four years), were reproduced rather successfully by the coupled CHM formulation for Eurobitum BW. A long-term model prediction of the evolution of the osmosis-induced pressure in the nearly constant volume tests shows that the pressure would reach a maximal value of about 20 MPa after about 5.5 years, after which the pressure would start to decrease. After 10,000 days (～27 years) the pressure would have decreased to a value of ～2 MPa.     

Availability of a probabilistic cost estimation for the price effect of Cu powder and bentonite on an HLW disposal cost in Korea

This study aims to demonstrate the availability of a probabilistic cost estimation related to the price effects of Cu powder and bentonite. From a sensitivity analysis of those materials’ prices on the overall disposal costs, it was found that Cu powder was a more dominant cost driver than that of bentonite among the material costs to dispose of 52,000 tU of spent fuels by the deterministic cost estimation method even though the used volume of Cu powder will be smaller than that of bentonite, whereas those conclusions can be changed by a probabilistic cost estimation method. Namely, its conclusion depends on a decision maker's personal opinion because of the resultant uncertainties. The disposal cost includes too many uncertainties due to the long construction and operational durations of a repository. Therefore a probabilistic cost estimation can be useful to provide the information related to an uncertainty.     

Processing highly tritiated water desorbed from molecular sieve bed using PERMCAT

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 10¹⁴ Bq kg^?1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.     

高放废物处置库缓冲材料中热力学过程分析

通过对高放废物深地层处置库缓冲材料中热力学过程的理论分析,建立起此缓冲层的物理模型和数学模型,并就所建模型的实用性和应用效果予以阐明。     

Design of FPGA based high-speed data acquisition and real-time data processing system on J-TEXT tokamak

Tokamak experiment requires high-speed data acquisition and processing systems. In traditional data acquisition system, the sampling rate, channel numbers and processing speed are limited by bus throughput and CPU speed. This paper presents a data acquisition and processing system based on FPGA. The data can be processed in real-time before it is passed to the CPU. It provides processing ability for more channels with higher sampling rates than the traditional data acquisition system while ensuring deterministic real-time performance. A working prototype is developed for the newly built polarimeter–interferometer diagnostic system on the Joint Texas Experimental Tokamak (J-TEXT). It provides 16 channels with 120 MHz maximum sampling rate and 16 bit resolution. The onboard FPGA is able to calculate the plasma electron density and Faraday rotation angel. A RAID 5 storage device is adopted providing 700 MB/s read–write speed to buffer the data to the hard disk continuously for better performance.     

Preparation,characterisation and dissolution of a CeO2 analogue for UO2 nuclear fuel

The behaviour of spent nuclear fuel under geological conditions is a major issue underpinning the safety case for final disposal. This work describes the preparation and characterisation of a non-radioactive UO₂ fuel analogue, CeO₂, to be used to investigate nuclear fuel dissolution under realistic repository conditions as part of a developing EU research programme. The densification behaviour of several cerium dioxide powders, derived from cerium oxalate, were investigated to aid the selection of a suitable powder for fabrication of fuel analogues for powder dissolution tests. CeO₂ powders prepared by calcination of cerium oxalate at 800 °C and sintering at 1700 °C gave samples with similar microstructure to UO₂ fuel and SIMFUEL. The suitability of the optimised synthesis route for dissolution was tested in a dissolution experiment conducted at 90 °C in 0.01 M HNO₃.     

Recent activities in the field of nuclear waste management

Buffer construction using bentonite pellets as filling material is a promising technology for enhancing the ease of repository operation. In this study, a test of such technology was conducted in a full-scale simulated disposal drift, using a filling system which utilizes a screw conveyor system. The simulated drift, which contained two dummy overpacks, was configured as a half-cross-section model with a height of 2.22m and a length of 6.0 m. The average dry density of the buffer obtained in the test was 1.29 Mg/m³, with an angle of repose of 35 to 40 degrees. These test results indicate that buffer construction using a screw conveyor system for pellet emplacement in a waste disposal drift is a promising technology for repositories for high level radioactive wastes.