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.     

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.     

Thermo-mechanical FE-analysis of residual stresses and stress redistribution in butt welding of a copper canister for spent nuclear fuel

The transient and residual temperature, stress and strain field present during electron beam welding of a plane copper end to a canister for spent nuclear fuel is calculated by the use of FEM. The subsequent stress redistribution is calculated up to 10,000 years. The canister consists of two concentric cylinders, an inner steel cylinder containing the spent nuclear fuel and an outer copper cylinder. It was found that the maximum plastic strain (plastic+creep) accumulated in the (possibly brittle) heat affected zone is ≈7%, which seems to be well below the reported ductility for the copper used.     

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.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

The fracture and aerosol release of impacted HLW-glasses due to accidental canister drops in a disposal site

For the disposal of HLW-canisters in a salt dome, two different accident scenarios have to be considered, canister drops in the reloading hall or in a borehole with drop heights of 10 m and 600 m, and reference drop velocities of 14 m/s and 80 m/s.The experimental program had two parts:

• - Laboratory scale drop tests with bare and canistered waste glass probes (scale: 1:10) to obtain basic data.

• - Full scale drop tests with inactive HLW-canisters, specified as planned for the German salt repository (H = 1.335 m, Ø = 0.43 m, weight: 550 kg, canister: SST 1.4833, wall: 5 mm).

The size distributions of the broken fines were measured by sieving and those of the filtered airborne particles by particle size analysis. The dominating parameter is the impact velocity (i.e. impact energy), further test parameters show no measurable influence, especially the canister influence on the fracture or aerosol release is negligible.Source terms, evaluated for the respirable fraction (particles with d < 10 μ m are between 2 × 10⁻⁴% for a 10 m drop and 0.1% for a 600 m borehole drop.     

300#反应堆乏燃料元件转运屏蔽吊筒的放射性屏蔽校核计算

介绍了根据300#堆乏燃料元件组件的实测剂量数据,对初步设计的乏燃料元件转运屏蔽吊筒的放射性屏蔽进行的详细校核计算。给出了乏燃料元件屏蔽前后不同距离处的剂量率。计算结果与实际验证表明屏蔽吊筒所选取的屏蔽厚度是合适的。     

Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea

Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl–NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.     

High-density support matrices: Key to the deep borehole disposal of spent nuclear fuel

Deep (4-5 km) boreholes are emerging as a safe, secure, environmentally sound and potentially cost-effective option for disposal of high-level radioactive wastes, including plutonium. One reason this option has not been widely accepted for spent fuel is because stacking the containers in a borehole could create load stresses threatening their integrity with potential for releasing highly mobile radionuclides like ¹²⁹I before the borehole is filled and sealed. This problem can be overcome by using novel high-density support matrices deployed as fine metal shot along with the containers. Temperature distributions in and around the disposal are modelled to show how decay heat from the fuel can melt the shot within weeks of disposal to give a dense liquid in which the containers are almost weightless. Finally, within a few decades, this liquid will cool and solidify, entombing the waste containers in a base metal sarcophagus sealed into the host rock.     

Multiple lines of evidence for performance of the canister and waste form in long-term nuclear waste disposal: Reviews

The time scales required for nuclear waste disposal are very large compared with those for other engineering endeavors. Because of this, there are many uncertainties associated with the quantitative performance assessment of canisters containing high-level radioactive waste in a waste form. Multiple lines of evidence can be helpful in building confidence in the long-term behavior (corrosion and dissolution) of the canister and waste form. These lines of evidence are derived from long-term supports and probabilistic models and developed based on shorter term tests, bounding and conservative approaches, and available observations on natural analogs. This paper presents the progress made for important lines of evidence considered in quantitatively assessing radionuclide release behavior from canisters and waste forms. This paper considers risk-significant issues for canisters and waste forms (i.e., risk informed approach) in the probabilistic performance assessment of the disposal system which has also other components such as geology and hydrology.     

MapGIS在东天山高放废物地质处置库选址中的应用

根据高放废物选址的要求,利用MapGIS对东天山地区不同时代岩体的空间位置、分布形态、出露面积,以及岩体与断裂构造、地震和矿床点等的空间关系进行了研究,对海量资料信息进行分析处理,可起到事倍功半的效果.利用MapGIS分析结果,初步筛选出阿奇山1号岩体、雅满苏北岩体作为高放废物地质处置库选址的有利岩体.     

高放废物地质处置元数据设计与编辑模块开发

在对元数据的含义、作用、组成及对高放领域研究的数据内容和特点分析的基础上,对元数据编辑模块的设计进行了详细描述,并在VS2005平台下应用C#语言进行了模块的实际开发,最终实现了便于进行数据共享的XML格式存储功能。元数据编辑模块的应用将促进我国高放废物地质处置信息化工作的进一步深入。     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

Development of monitoring technique for the confirmation of spent fuel integrity during storage

A method is developed to monitor integrity of spent fuels stored in a canister that is sealed by weld. To achieve the monitoring, Kr-85 gas was newly adopted as a kind of probe. In the case of a fuel rod defect, Kr-85 gas of the fuel rod is leaked in the canister. By detection of gamma ray (514 keV) emitted from Kr-85 outside of the canister, defected rods can be detected without unsealing the canister. The monitoring technique was developed using small-scaled mock-up experiments and simulated calculation. The result showed that Kr-85 gas leakage of about 10¹¹ Bq is detectable under the noise gamma rays by using the detection system with collimator, which is about 10% of Kr-85 inventory in a fuel rod. Therefore, this monitoring technique is considered as an inspection method prior to transportation of spent fuel from an interim storage facility to a reprocessing plant.     

Study of an optimization approach for a disposal tunnel layout, taking into account the geological environment with spatially heterogeneous characteristics

The geological environment has spatially heterogeneous characteristics with varied host rock types, fractures and so on. In this case the generic disposal tunnel layout, which has been designed by JNC, is not the most suitable for HLW disposal in Japan. The existence of spatially heterogeneous characteristics means that in the repository region there exist sub-regions that are more favourable from the perspective of long-term safety and ones that are less favourable. In order that the spatially heterogeneous environment itself may be utilized most effectively as a natural barrier system, an alternative design of disposal tunnel layout is required. Focusing on the geological environment with spatially heterogeneous characteristics, the authors have developed an alternative design of disposal tunnel layout. The alternative design adopts an optimization approach using a variable disposal tunnel layout. The optimization approach minimizes the number of locations where major water-conducting fractures are intersected, and maximizes the number of emplacement locations for waste packages. This paper will outline the variable disposal tunnel layout and its applicability.     

高放废物模拟地质处置研究平台建立、运行及初期成果

高放废物的妥善处置是核能可持续发展的前提,在国际范围内受到高度重视,地质处置是普遍接受的方案.高放废物模拟地质处置研究平台由25套模拟多重屏障系统及其共用的低氧环境构成,可以模拟多种不同处置条件下的核素浸出情况.完成部分系统的装填运行.结果表明,多种元素的浸出受到了多重屏障的抑制,并随着时间的推移浸出浓度趋于稳定;不同的处置温度、玻璃体类型、围岩类型、膨润土含素玻璃粉均对重要核素的浸出有显著影响;多种包装材料的耐蚀性能差异显著.下一步实验中,根据现有的研究结果,对玻璃体、膨润土、包装材料等的装填进行优化,对取样系统进行改进,完善总体的实验方案,研究多种处置条件对元素的浸出影响.     

Application of a remote-controlled collimated dosimetric system for surveying spent MR fuel storage facilities

The results of using a specially developed collimated dosimetric system, placed on a Brokk-90 remotecontrolled machine, to survey a spent fuel storage facility are presented. The measuring block of the system consists of collimated and uncollimated γ-ray detectors, video cameras, and headlights. The range of the dose-rate measurements of the γ-ray detectors varies from 0.4 mSv/h to 8.5 Sv/h. The collimation angle of the collimated detector is 12° and the lead shielding is 30 mm thick. The system made it possible to perform this work in large radiation fields by remote means. The dose rate distribution along extended elements, extracted from the spent fuel storage facility in the MR hall, is obtained. The contents of the storage facility were visualized in all details. Using the results of the survey, the contents of the storage facility were repacked and removed from the MR room.