Feasibility Study on Transportation Techniques Employing Air-Bearing System for Pre-Assembled Waste Package at Drift Tunnel of HLW Repository

Full-scale tests were performed to evaluate the technical feasibility of a transport system with air-bearings at the underground HLW disposal tunnel for pre-assembled heavy disposal packages, which consist of a waste package and buffer material. Transport conditions in the disposal tunnel, such as roughness and unevenness of the curved surface, make it difficult to achieve smooth movement using the commercial airbearing transport system. In order to evaluate the applicability of the air-bearing transport system to such conditions, tests using a full-scale test device (modified package) and simulated tunnel surface were conducted. Based on the tests, the applicability of this transport system to a disposal tunnel was confirmed.     

Preliminary conceptual design of a geological disposal system for high-level wastes from the pyroprocessing of PWR spent fuels

The inventories of spent fuels are linearly dependent on the production of electricity generated by nuclear energy. Pyroprocessing of PWR spent fuels is one of promising technologies which can reduce the volume of spent fuels remarkably. The properties of high-level wastes from the pyroprocessing are totally different from those of spent fuels. A geological disposal system is proposed for the high-level wastes from pyroprocessing of spent fuels. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. Around 665 kg of monazite ceramic wastes are expected from the pyroprocessing of 10 MtU of PWR spent fuels. Decay heat from monazite ceramic wastes is calculated using the ORIGEN-ARP program. Disposal modules consisting of storage cans, overpacks, and a deposition hole or a disposal tunnel are proposed. Four kinds of deposition methods are proposed. Thermal design is carried out with ABAQUS program and geological data obtained from the KAERI Underground Research Tunnel. Through the thermal analysis, the spacing between the disposal modules is determined for the peak temperature in buffer not to exceed 100 °C. Thermal analysis shows that the optimum spacing between the vertical deposition holes with 4 overpacks is 8 m when the disposal tunnel spacing is 40 m and optimum spacing of 2 m for horizontal disposal tunnel with 25 m tunnel spacing. Also, the spacing reduces to 6 m for vertical deposition when the double-layered buffer is used, which reduces the disposal area to one-sixty fifth (1/65th) compared with the direct disposal of spent fuels. Finally, the effect of cooling time on the disposal area is illustrated.     

钠雾化喷射技术研究

钠雾化喷射技术是大量放射性废钠处理的一项关键技术,为探索钠雾化工艺,设计了钠雾化设备,并对钠雾化喷射的动力学性能进行研究。试验在体积为2.4 m³的密封容器内进行。氩气气氛下,采用激光粒度仪测量距喷头150 mm处钠喷射液滴粒径的分布,试验压力分别为0.05、0.1和0.15 MPa。试验结果表明：3种压力下,喷射中位粒径分别为567.2、554.8和544.2 μm,中位粒径随喷射压力的变大而变小;不同压力下,钠喷射液滴均在464.2～735.6 μm区间的粒径分布最多,占总分布的50%以上。通过钠喷射与水喷射的相似性,建立粒径相似性模型,利用已知水喷射液滴粒径空间分布模拟得出钠喷射液滴粒径的空间分布。     

CFD技术与SACTI模型在自然通风冷却塔雾羽局地环境影响模拟中的应用比较研究

本文利用CFD技术和SACTI模型对山东某电厂冷却塔附近流场以及冷却塔雾羽抬升以及漂滴与盐的沉积对局地环境影响进行了模拟计算,并与现场试验结果进行对比分析验证。结果表明,二者模拟结果与现场试验数据基本吻合, SACTI模型预测漂滴与盐沉积的最大距离出现在冷却塔附近200 m范围内,CFD技术模拟漂滴与盐沉积的最大距离出现在下风向500 m左右。SACTI模型预测漂滴与盐地面最大沉积量约为CFD技术模拟地面最大沉积量的3倍,主要原因是由于CFD在模拟过程中考虑的冷却塔局地环流以及周围建筑物对漂滴沉积的影响。     

近场核素释放率对缓冲材料参数的敏感性研究

缓冲材料对高水平放射性废物(高放废物)地质处置库的安全至关重要。本文在处置库关闭后预期演变情景分析的基础上,运用蒙特卡罗随机模拟方法,对缓冲层厚度、缓冲材料密度、核素在缓冲材料中的分配系数这三个参数进行敏感性分析。结果表明,处置库关闭后1 000 a内,近场核素释放率对缓冲层厚度较敏感,超过1 000 a 后敏感性较低;近场核素释放率对缓冲材料密度不敏感;核素在缓冲材料中的分配系数不断提升的情况下,对应的参数敏感度也逐渐加大。上述敏感性分析结果可为缓冲材料工程设计提供参考。     

Iodine release behavior from iodine-immobilized cement solid under geological disposal conditions

The iodine release behavior from the iodine-immobilized cement was investigated. From the results of immersion tests using ion-exchanged water (IEW) and calculations, the solubility equilibrium model could describe the iodine release behavior. To assess the performance of cement in an actual environment, it is important to confirm that the solubility equilibrium model is applicable to the geological disposal conditions. From immersion tests using simulated ground water, the release of iodine from the cement occurred in a shorter period of time than in the case of IEW, and reactions with CO₃²⁻ and Cl⁻, which were contained in the simulated ground water, had an influence on the iodine release behavior. As a result of calculations using the solubility equilibrium model, the liquid-solid ratio at which the iodine was completely released was mostly in agreement with the results of immersion tests. The results show that this model can be applied in a wide range of environments.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

Integrity Test of Multi-Stage Design Packages of Radioactive Wastes under Deepsea Condition

For sea disposal of the low-level radioactive wastes, high hydrostatic pressure tests on the full size (200 l) multi-stage type packages were carried out in a pressure vessel. Using the data obtained, ingress of water through leak path was simulated by a computer analysis.

In order to confirm the above results, a demonstration test on integrity of the package in deepsea (5,000 m depth) was carried out at 90 miles off Nojimazaki, Chiba-ken (143° 10°E, 33°507°) by hanging the package down to 5,000 m depth. In these tests, no appreciable damage of the packages was observed which could give rise to controversy in safety.     

Separation Technology for Radioactive Iodine from Off-Gas Streams of Nuclear Facilities

Iodine separation technology using an inorganic adsorption material has been investigated in order to apply the technology to the off-gas treatment systems of nuclear facilities. Iodine removal efficiencies were checked by laboratory experiments using simulated off-gas streams of various conditions and the developed adsorbent, silver impregnated alumina (AgA). Laboratory test results demonstrated effective iodine removal with high decontamination factors (DF's) at relatively high temperatures (≥100°C). Then the removal efficiency were confirmed using actual off-gas streams sampled from the dissolver off-gas treatment system of the Karlsruhe reprocessing plant. The DF's were over 10³ with the AgA bed depth of 10 cm and showed little change during the adsorption period, which indicated applicability of the iodine removal technology with AgA to nuclear fuel reprocessing plants. Iodine adsorption capacity and its release property were also investigated using simulated off-gas streams. The former had a value of ~0.22g/g-AgA and this value could well predict the breakthrough property. The adsorbed iodine was judged stable during the storage of AgA saturated with iodine in air at temperatures below 500°C and in water at ~20°C after changing the adsorbed iodine form from AgIO₃ to Agl. Thus, the separation technology provided effective and stable iodine separation from the off-gas of nuclear facilities.     

Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton^® GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 °C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton^® A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large diameter, high temperature seals of PFBR are indicated along with the envisaged activities en-route the production of perfected reactor inflatable seals.     

Safety study on HIC containing waste resin with respect to hydrogen release

To explore the behavior of radiolytically produced hydrogen release from the waste resin stored in a high integrated container(HIC), and the mechanism of hydrogen diffusion in a near-surface disposal facility, both experimental studies and numerical simulations were performed through an accelerated irradiation test and simulated disposal, respectively. Results indicated that,100 years after disposal, the highest hydrogen concentration appeared in the cell where the HICs were placed. The volume fraction for different scenarios postulated in the numerical simulation was 2.64% for Scenario 1, 2.28% for Scenario 2, and 3.965% for Scenario 3, all of which are lower than the hydrogen explosion limit of 4.1%. The results indicated that the simulated HIC disposal scheme is safe.     

CEFR一回路冷阱更换过程中的钠钾合金灌装

介绍了中国实验快堆(CEFR)一回路2#冷阱更换过程中的钠钾合金灌装方案,灌装回路的设计、建造及调试,灌装过程安全防护措施和废钠钾合金的处理,并向CEFR备用冷阱内灌装447.4kg钠钾合金,对废钠钾合金进行处理。结果表明,所采用的钠钾合金灌装方法、钠钾合金回路活接头冷冻拆卸技术、钠管道切割除钠技术、废钠钾合金处理方法安全可靠,人员安全防护措施得当。本文方法也可用于指导CEFR一回路1#冷阱的更换。     

Physical properties and leaching behaviour of spent fuel BISO coated particles

The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The behavior of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size and morphological characteristics of the coated particles were investigated by the using of optical and SEM microscopy. The distribution of the ¹³⁷Cs amount in the coated particle was studied in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total ¹³⁷Cs only. Further, the thoria-based (Th_0.834U_0.166)O₂ kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. In all experiments the average flow rate was ∼7–8 ml/day. Dissolution of irradiated and unirradiated kernels in HCl solution with the different value of pH (from 0 to 5) was investigated at the temperatures 90, 55 and 20 °C. The amounts of the radionuclide leached in solutions were determined by ACP-MS, γ- und α-spectrometry. On the basis of the obtained results the important leaching characteristics such as the normalized leaching rate, the activation energy value for the release of the different radionuclides were calculated.     

Numerical analysis of highly efficient laser-based method of radioactive Cs isotope separation utilizing light-induced drift in D1 and D2 transitions in rare gases

We investigated the feasibility of separating radioactive Cs isotopes using the effects of light-induced drift (LID). LID is the massive flow of particles caused by the difference between the collision frequencies of the buffer particles in the optically connected ground and excited states. We numerically evaluated the LID velocities of Cs-133, Cs-135, and Cs-137 in rare gases based on the currently available experimental technologies. The calculations were performed utilizing the steady-state analytical model and assuming that all of the collisions were strong. The velocity-dependent collision cross-sections were estimated using the WKB approximation with the recently developed ab initio atom–atom interaction potentials. A maximum velocity of more than 10 m/s was obtained when He was utilized as the buffer gas, and the expected maximum annual throughput was approximately 600 g, which is sufficient for practical purposes. We also examined the negative effects that were neglected in the employed model.     

BESⅢ主漂移室内室升级MAPS芯片探针台测试系统设计

为满足北京谱仪Ⅲ主漂移室内室升级MAPS芯片测试需要,设计了一套用于芯片功能检查的芯片探针台测试系统。该系统实现了批量芯片JTAG通讯、芯片功耗、像素箝位电压、读出数据等芯片的功能检查,并可以进行芯片噪声水平以及甄别器阈值扫描等初步测试。芯片的探针台测试不仅在非邦定的情况下完成了芯片的筛选,同时可为后续芯片在探测器上工作时的阈值等参数配置提供参考。     

Double-layered buffer to enhance the thermal performance in a high-level radioactive waste disposal system

A thermal performance is one of the most important factors in the design of a geological disposal system for high-level radioactive wastes. According to the conceptual design of the Korean Reference disposal System, the maximum temperature of its buffer with a domestic Ca-bentonite is close to the thermal criterion, 100 °C. In order to improve the thermal conductivity of its buffer, several kinds of additives are compared. Among the additives, graphite shows the best result in that the thermal conductivity of the bentonite block is more than 2.0 W/m °C. We introduced the concept of a double-layered buffer instead of a traditional bentonite block in order to use the applied additive more effectively. The thermal analysis, based upon the three-dimensional finite element method, shows that a double-layered buffer could reduce the maximum temperature on a canister's surface by 7 °C under identical conditions when compared with a single-layered buffer. An analytical solution was derived to efficiently analyze the effects of a double-layered buffer. The illustrative cases show that the temperature differences due to a double-layered buffer depend on the thickness of the buffer.     

A Thermoelectric-Conversion Power Supply System Using a Strontium Heat Source of High-Level Radioactive Nuclear Waste

A thermoelectric-conversion power supply system with radioactive strontium in high-level radioactive waste has been proposed. A combination of Alkali Metal Thermo-Electric Conversion (AMTEC) and a strontium fluoride heat source can provide a compact and long-lived power supply system. A heat source design with strontium fluoride pin bundles with Hastelloy cladding and intermediate copper has been proposed. This design has taken heat transportation into consideration, and, in this regard, the feasibility has been confirmed by a three-dimensional thermal analysis using Star-CD code. This power supply system with an electric output of 1MW can be arranged in a space of 50m² and approximately 1.1m height and can be operated for 15 years without refueling. This compact and long-lived power supply is suitable for powering sources for remote places and middle-sized ships. From the viewpoint of geological disposal of high-level waste, the proposed power supply system provides a financial base for strontiumcesium partitioning. That is, a combination of minor-actinide recycling and strontium-cesium partitioning can eliminate a large part of decay heat in high-level waste and thus can save much space for geological disposal.     

射线检测数字化成像技术在核电施工中应用可行性的研究

射线检测为核电施工中最主要的无损检测方法之一,每台机组建造过程中大概需要拍摄二十万套底片,而传统胶片照相检测技术存在诸多不足,故迫切需要寻求一种全新高质量的射线检测方法。数字化成像CR技术有希望替代胶片进行射线检测的技术,选定CR技术做为核电施工中运用可行性分析研究的主体,利用研发制造的验证性测试先进试块,对CR图像质量影响因素和CR技术特征进行测试研究。建立并使用庞大的缺陷试样库,制定并完善可运用于核电现场管道焊缝对比测试的工艺参数,实施对核电现场各系统管道焊缝CR技术和胶片技术的对比透照,进而从CR技术特点、技术可行性、应用经济性、操作适应性和便利性等角度对CR技术在核电施工中的应用进行可行性分析。     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

宇宙射线μ子探测高Z材料的仿真研究

利用Geant4建立宇宙射线μ子探测高Z材料的仿真系统,采用最大似然算法,研究用于μ子探测的位置灵敏探测器漂移管的位置分辨率及系统成像时间对成像结果的影响。当漂移管的位置分辨率优于200μm、成像时间在1～2min内可对相对独立的高Z物质快速识别,本研究也为漂移管的设计加工提供了理论依据。