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.     

Concept of a Korean Reference Disposal System for Spent Fuels

A deep geologic disposal system for the spent fuels from nuclear power plants has been developed since this program was launched in 1997 in Korea. In this paper, the concept of a Korean reference high-level waste (HLW) vertical disposal system (KRS-V1) is described. Though no site for the underground repository has yet been specified in Korea, a generic site with a granitic rock is considered for a reference spent fuel repository design. The depth of the repository is assumed to be 500 m. The repository consists of a disposal area, a controlled area, and an uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. In the controlled area and the uncontrolled area, there are technical rooms and tunnels and/or shafts to connect them to the ground level, respectively. The repository will be excavated, operated, and backfilled in several phases including an underground research laboratory (URL) phase. The result of this concept development will be used for an evaluation of its feasibility, analyses of its long-term safety, information for public communication, and a cost estimation, among others.     

The role of pyro-processing in decreasing disposal cost in Korea

A comparative analysis regarding the disposal cost of HLW (High-Level Waste) from 20,000-ton PWR nuclear fuel, focusing on pyro-processing and direct disposal, was conducted in this study. A cost estimation of the major cost drivers in disposing of pyro-processed waste revealed that canisters would cost 67.32 MEUR and that the disposal holes and disposal tunnels would require about 11.2 MEUR for excavation. These estimates amount to 1/16 and 1/55 of the costs for direct disposal of PWR spent fuels, respectively. These significant disposal cost savings in pyro-processed radioactive waste result from a significant reduction in the amount of radioactive waste to be disposed of thanks to the recycling in a fast reactor.     

Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

A thermal model is constructed and analyses are performed for an ‘in-floor’ type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels.     

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.     

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.     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

高放废物地质处置黏土岩处置库围岩研究现状

世界上很多国家都对处置库的可能围岩进行了详细研究。通过对比,认为花岗岩、黏土岩、岩盐比较适合作为处置库围岩,而黏土岩由于具有自封闭性、渗透率低等其他岩石类型不可比拟的优点,因而将黏土岩作为高放废物地质处置库围岩越来越受到各国的关注。文章同时介绍了瑞士、法国、比利时等国家在黏土岩中所进行的大量研究,均认为在黏土岩中处置高放废物和乏燃料是安全的。文章还对黏土岩处置库概念设计、黏土岩处置库围岩地下实验室研究,以及我国开展黏土岩处置库研究的意义等进行了综述。     

高放废物处置库围岩开挖损伤区研究的关键问题及进展

高放废物(HLW)地质处置是将高水平放射性废物埋存于地下500～1 000 m地质体中,使放射性废物与生物圈长期隔离。地质处置库对核素的长期隔离能力是安全评价的关键课题。地下硐室的开挖将不可避免地对围岩造成损伤,形成开挖损伤区(EDZ),改变围岩的物理力学特性,对高放废物地质处置长期安全性存在潜在的影响。目前多个国家建成了高放废物处置地下实验室,并开展了大型原位开挖损伤区的研究,研究开挖损伤区的形成过程及其物理力学特性的变化。本文综述了国外结晶岩地下实验室开展的开挖损伤区研究,总结了EDZ关键研究问题;梳理了加拿大、瑞典、芬兰3个地下实验室多年来开展的系统的EDZ研究工作,对当前EDZ预测模型及模拟技术进行了总结;对我国地下实验室将开展的开挖损伤区研究工作进行了初步探讨,期望为我国的相关研究提供借鉴。同时,高放废物处置库是地下工程新实践,其EDZ的研究成果,形成的技术方法将对其他行业地下工程的建设,如引水隧洞、公路铁路隧道等也有重要的参考价值。     

高放废液中锕系离子分离研究进展 Ⅰ.双酰胺荚醚与锕系离子的配位化学

如何处理处置核电站反应堆产生的乏燃料及乏燃料后处理过程产生的高放废液是发展安全核能面临的一个主要问题。为提高核能的安全性、减少需要长时间深地层处置的高放废物量、有效利用地球上有限的可裂变材料资源,世界上发展核能的国家在过去几十年发展了从高放废液中分离少量锕系元素离子的萃取分离流程。近年来,双酰胺荚醚类化合物在锕系元素分离方面备受关注,本文从基础配位化学角度综述近期这类化合物与锕系元素离子相互作用等方面的研究结果。     

Review of the development of the transportation,aging, and disposal (TAD) waste disposal system for the proposed Yucca Mountain geologic repository

The U.S. Department of Energy (DOE) began studying Yucca Mountain in 1978 to determine whether it would be suitable for the nation’s first long-tem geologic repository for over 70,000 metric tons of spent (or used) nuclear fuel and high-level radioactive waste. The purpose of the continuing Yucca Mountain study, or project, is to comply with the Nuclear Waste Policy Act of 1982 as amended in 1987 and develop a national disposal site for spent nuclear fuel and high-level radioactive waste disposal. In 2005, DOE shifted the design of the proposed repository from a concept of unloading spent nuclear fuel from transportation canisters and loading into disposal canisters (which required a great deal of handling radioactive material at the repository site) to a “clean” facility, unveiling the transportation, aging, and disposal (TAD) canister system. The TAD waste system consists of a canister loaded with commercial spent nuclear fuel.This review paper provides a comprehensive review on the status of TAD, technical and licensing requirements, the work that has been done so far, and the challenges and issues that must be addressed before TAD can be successfully implemented. Though the future of the Yucca Mountain project is bleak at this point, the progress that has come in the field of TAD will be one of its lasting legacies.     

高放废物处置库近场裂隙水流-传热-处置室间距相互作用的三维数值分析

为研究高放废物地质处置库近场裂隙水流-传热-处置室间距的相互作用机理,采用3DEC软件计算裂隙水流-传热-处置室间距相互作用对处置库近场温度分布影响。结果表明:(1)在处置室间距相同条件下,流动的裂隙水显著改变了处置库近场温度场,使岩体温度降低,缩短模型达到稳态所需要的时间。(2)处置室间距增大,温度叠加效应减弱,处置库近场温度越低,并且废物罐表面膨润土温度越低,裂隙出水口水温越低,模型达到稳态所需要的时间越短。(3)水平和垂直裂隙水流共同传热使处置库近场裂隙水流下游区域温度显著高于裂隙水流上游区域。(4)处置室间距为6 m和8 m时,水平裂隙出水口水温高于垂直裂隙,处置室间距为10 m时,水平裂隙出水口水温低于垂直裂隙。     

Corrosion issues in nuclear waste disposal

This paper summarized some corrosion issues specific to nuclear waste disposal and illustrates them by the French geological clay concept for the reliable prediction of container degradation rate and engineering barrier integrity over extended periods, up to several thousands years. Among the items, the following are included:

• The importance of the underground repository conditions.

• The necessity of developing comprehensive semi-empirical models and also predictive models that must be based on the mechanisms of corrosion phenomena.

• The use of archaeological artefacts to demonstrate the feasibility of long term storage and to provide a database for testing and validating the models.

Long-term lining performance - Civil engineering problem of potential retrieval of buried spent nuclear fuel

The current solution for the spent fuel, high-level and long-lived radioactive waste is to store them at surface facilities from which they will be subsequently moved to a deep repository. No such repositories are in operation currently but several such facilities are close to the construction phase. A deep repository can be situated in several types of geological conditions including clay formations, salt sediments, argillites and tuffitic and granitic rocks. The character of the host rock is the key factor determining the design and specific requirements of individual components of such a facility. The future potential retrieval of canisters containing nuclear waste from the repository is a further influential factor. The reason for retrieval of containers lies in the development of fast reactors and increased interest for spent fuel reprocessing. Naturally, the decision as to whether retrievability is technically feasible must be made before finalising the design and construction process of the repository. If the decision is made to retrieve, a design which will include all the relevant safety aspects for the potential retrieval of canisters must be determined. The lay-out of the repository, the materials to be used and the design of the various structures of the facility (e.g. access tunnels, disposal shafts, buffer and backfill) are not the only issues to be addressed. The long-term stability of the system as a whole, i.e. of all the components, is crucial. Depending on the disposal concept chosen, the thermal load generated by the waste in the disposal container, saturation by water from the surrounding environment and the loading of the host rock massif will constitute the main processes which will affect the behaviour, safety and future functioning of the repository from the civil engineering point of view. The long-term stability of the lining of disposal galleries is a basic precondition for the safe removal of spent nuclear waste from deep underground repositories. The stability problems of tunnel linings exposed to long-term thermal load have not yet been properly addressed and form the subject of the European TIMODAZ project (Thermal Impact on the Damaged Zone around a Radioactive Waste Disposal in Clay Host Rocks) and also supported by the “Complex System of Methods for Directed Design and Assessment of Functional Properties of Building Materials” project. This paper describes the design, construction and currently available results of a 1:1 scale “in situ” disposal tunnel model which has been built at the Josef Underground Educational Facility in the Czech Republic.     

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。     

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。     

The economic effects of the deferred disposal of spent fuel in Korea

The present study analyzes the economic effects concerning deferred disposal of spent fuel through long-term storage. According to the cost analysis, a scenario that a 90-year deferral of an HLW (High-Level Waste) repository construction in favor of a long-term storage of spent fuel would be economically preferable to another scenario based on the year 2040 chosen as the starting point for construction on a repository. That is, the former scenario would cost about 1/2 of the latter. This finding is an estimated result from an economic perspective only, assuming the disposal of 20,000-ton PWR spent fuel and 16,000-ton CANDU spent fuel. Still, it seems necessary to elicit proper term of storage for radioactive waste in order to comply with the so-called Polluter-Pays principle that the current generation cannot pass on its radioactive waste to the next generation.     

Progress towards a Deep Repository for Spent Nuclear Fuel in Sweden

The Swedish Nuclear Fuel and Waste Management Co. has in operation a safe and well integrated system for handling of all radioactive residues within Sweden. The existing central repository for low- and medium-level waste (SFR) and the central interim-storage facility for spent nuclear fuel (CLAB) can accommodate all the radioactive waste produced inside Sweden. Comprehensive research, development and demonstration activities are well under way for an encapsulation plant and a deep repository for spent fuel. These two facilities remain to be constructed to complete the waste management system. Siting of the deep repository is in progress with the aim of finding a suitable and accepted site. Implementation of the deep geological repository is a technical, scientific, social and political challenge. Smooth implementation must take into consideration both facts and emotions. Patience, flexibility and respect for the democratic process are important keywords. Research facilities, such as the underground Äspö Hard Rock Laboratory and the Encapsulation Laboratory, are important to promote scientific understanding as well as to demonstrate the disposal concept and technology.     

高放废物处置的几个问题

就与高放废物处置有关的几个问题,如地下实验室进行场所问题,地下实验室与处置库建设安排问题,高放废物深地质处置的替代方法等进行了探讨,希望有助于我国高放废物处置工作的进一步开展。     

Development of packaging to transport new standardised range of disposal canisters to geological disposal facility in UK

Radioactive Waste Management Limited (RWM) of the Nuclear Decommissioning Authority (NDA) is developing concepts to demonstrate the viability of using a standardised range of disposal canister (DC) designs for geological disposal of high level waste and spent fuel in the UK. The standardised DC are designed for disposal in a geological disposal facility with integrity requirements in the range 10?000 to 100?000 years. International Nuclear Services (INS) is also a subsidiary of the NDA and working with RWM to develop a design of packaging for transporting these DC, which is called the disposal canister transport container (DCTC). Initial studies undertaken by INS focused on optimising payload and geometry for the canister designs. Subsequent studies focused on achieving criticality safety requirements for transport, which established the use of multiple water barriers, were required for higher enriched spent fuels. The results of this initial work were presented at the International Nuclear Engineering society conference at London in 2012. Subsequently, RWM commissioned INS to develop the design of DCTC to a level where it would be viable for licensing as a transport package with appropriate level of technical understanding. A specific requirement of RWM was that the loaded DCTC should be capable of transportation on an existing design of four axle rail wagon, within a gross mass of 90 t, this giving considerable logistic and overall cost benefits. Recent development work has focused on detailed impact, thermal and shielding analysis and how these influence the DCTC transport mass and the position of that mass in relation to the four axle rail wagon, both of which influence its capability for the required transport. In terms of meeting mass limits, achieving the specified radiation shielding performance (neutron and gamma) for the spent fuel was found to be a major challenge. However, of equal challenge was to accommodate the high forces generated under impact accident conditions due to the high mass ratio of contents to container. In order to mitigate these forces, the shock absorber designs needed to be carefully judged because their dimensions were restricted by the rail wagon design. This paper describes the DCTC development work, how the design challenges were addressed and the conclusions reached.