Management of radioactive wastes from the operation of nuclear power plants

A prerequisite for the acceptance of the nuclear energy system is the effective management of the rad-wastes. Among the wastes to be considered, there are the wastes from the operation and decommissioning of nuclear power plants, as well as those from the nuclear fuel cycle. For the management of operating wastes, processes and facilities optimized in the course of several decades, are available, with which the raw solid and liquid wastes can be reduced in volume and turned into products which are physically and chemically stable and thus suitable for final disposal. The management of spent fuel can be done either by direct final disposal or reprocessing. The required interim storage facilities are ready for operation. The methods and a facility for packaging spent fuel for direct final disposal are in an advanced stage of development and construction. If fuel assemblies are to be reprocessed abroad, the wastes generated from the process must be taken back. Decommissioning wastes have technical properties which correspond essentially to the various groups of operating wastes and can thus be processed with similar methods; however since large quantities of them are generated in relatively short times, they present particular logistic problems. All waste types end up in final disposal sites to be built under the responsibility of the federal government. A final disposal site for low level wastes is in operation. In addition, two final disposal projects for accommodating higher level wastes including spent fuel for direct disposal and vitrified wastes from reprocessing, are being pursued.     

Radioactive waste management practices in India

Concern for the environment and establishment of radiation protection goals have been among the major priorities in planning of India's nuclear energy programme. In the Indian nuclear fuel cycle, right from inception, a closed loop option has been adopted where spent fuel is reprocessed to recover plutonium and unused uranium. The emphasis has been to recover actinides, individual fission products and recycle them back to the fuel cycle or use them for various industrial applications. The development of innovative treatment processes for low and intermediate level wastes in recent times has focused on volume reduction as one of the main objectives. In the case of high-level liquid waste, vitrification in borosilicate matrix is being practiced using induction heated metallic melters at industrial scale plants at Tarapur and Trombay.Currently, there are seven operating near surface disposal facilities co-located with power/research reactors in various parts of the country for disposal of low and intermediate level solid wastes. These are routinely subjected to monitoring and safety/performance assessment. An interim storage facility is operational for the storage of vitrified high-level waste overpacks for 30 years or more. Nation wide screening of potential regions and evaluation of rock mass characteristics is in progress for ongoing geological repository programme. Preliminary design and layout of an underground research laboratory/repository has also been initiated.A research programme is underway for long-term evaluation of vitrified waste product under simulated repository conditions. Research is also directed towards development of advanced technologies for waste processing as well as conditioning in vitreous and ceramic matrices. The Department of Atomic Energy with participation of the Indian industry has developed all essential remote-handling gadgets required for operation and maintenance of waste management system and assemblies including decommissioning.     

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.     

The Belgium underground research facility: Status on the demonstration issues for radioactive waste disposal in clay

Safe disposal of radioactive waste is one of the key issues in the consolidation process of the nuclear industry. Disposal in deep geological formations is at present the most promising option for long-lived waste and spent fuel. In Belgium, from a survey of potential geological formations, preference was expressed for a deep clay formation (Boom clay) lying under the facilities of the Mol Research Centre. This does not anticipate at all the site selection procedure. Rather early in the programme (1980) an underground laboratory was constructed in the clay to evaluate feasibility aspects and to become an in situ facility for performing tests in close to real conditions within the HADES (High Activity Disposal Experimental Site) project. The present programme requires now a major extension of the present underground facility to perform new large-scale in situ tests. A major action to be developed for the next 10 years is the PRACLAY project, a demonstration experiment simulating the thermal output of a 30 m long high level waste disposal gallery, 2 m in diameter. The experiment will be installed from an extension of the existing facilities to be built over the next 3 years. The experiment is planned to last until 2005.     

Durability of a reinforced concrete designed for the construction of an intermediate-level radioactive waste disposal facility

The National Atomic Energy Commission of the Argentine Republic is developing a nuclear waste disposal management programme that contemplates the design and construction of a facility for the final disposal of intermediate-level radioactive wastes. The repository is based on the use of multiple, independent and redundant barriers. The major components are made in reinforced concrete so, the durability of these structures is an important aspect for the facility integrity. This work presents an investigation performed on a reinforced concrete specifically designed for this purpose, to predict the service life of the intermediate level radioactive waste disposal facility from data obtained with several techniques. Results obtained with corrosion sensors embedded in a concrete prototype are also included. The information obtained will be used for the final design of the facility in order to guarantee a service life more or equal than the foreseen durability for this type of facilities.     

Challenges in spent nuclear fuel final disposal： conceptual design models

The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.     

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.     

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.     

Cooldown testing of nuclear fuel casks

Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown.     

Thermal characteristics of aged spent fuel in a dry environment

Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown.     

Waste Package and Shipment Aspects in Relation to Transport and Disposal Requirements

Abstract

In the management of radioactive waste, different processes have to be considered such as conditioning, interim storage and final disposal together with transport as the linking process. Attention should be paid to all the relevant steps within these processes, in particular to derive appropriate waste package requirements for a safe waste management system as well as to obtain a consistent regulatory framework. Radioactive waste arising from research and development centres, nuclear power plant operation, decommissioning, the nuclear fuel cycle industry, and applications of radioisotopes in medicine, industry and research, has finally to be shipped to a final disposal site. Therefore waste packages are subject to both the regulatory requirements of transport and the requirements of disposal. Resulting consequences for waste package limitations will be discussed, in particular for low and intermediate level waste taking into account LSA/SCO regulations for transport and waste acceptance criteria for disposal in Germany. Some aspects of different package concepts, like the use of non-reusable or reusable packages, will be considered as well as the application of LSAISCO regulations and further development of LSA/SCO criteria.     

美国放射性废物最终处置政策及技术路线

反应堆安全和核废物安全处置被认为是影响今后核能事业发展的两大障碍。自１９８０年以来，美国颁布了３个关于核废物处置的政策法令，对放射性废物的安全处置的要求及责任做出了明确规定。文章介绍了美国放射性废物处置的政策、技术路线及现状。对乏燃料及高放废物的处置，美国采取了比较慎重的态度，进行了各种方案的比较，虽然已有基本轮廓，但仍在探索之中。文章还介绍了高放废物处置中存在的一些有争议的重大问题和倾向性意见。     

Storage and final disposal of spent htr fuel in the federal republic of Germany

The spent fuel treatment concept for HTR in the FRG is based on direct disposal of the fuel in a salt dome repository. Due to high burnup and good in situ fuel utilization, direct disposal offers economic advantages, especially for low enriched uranium fuel. Besides, the safety requirements can be met by simple techniques due to the special features of the HTR fuel element: coated particle fuel, stabilized in a graphite matrix with absence of any metal and the low heat production per volume unit give favourable preconditions for intermediate storage and safe disposal. Techniques for the intermediate storage are available and practised with AVR and THTR fuel. For final disposal, emplacement in boreholes, 300–600 m in depth, using simple packaging, was chosen as reference, similar to the reference concept for heat generating, medium-active waste. So far, the results of both the development and the experimental test programme underline the chosen concept.     

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.     

Challenges in the renewal of the Swedish sea transport system

Abstract

Since 1985, SKB has successfully operated a sea transport system for transport of spent nuclear fuel and radioactive waste to the intermediate storage facility, Clab and the final repository, SFR, in Sweden. The main components in the system are the ship M/S Sigyn, transport casks for spent fuel and core components, IP2 containers and terminal vehicles.     

乏燃料处置的必要性及其处置库环境化学行为

为了确保核燃料循环的安全性,不宜处理的乏燃料也应该同玻璃固化体一样作为高放废物进行深地质处置。本文综述了一些前期工作,归纳了空气侵入和水的辐解产生氧化性产物是导致乏燃料UO2基体氧化溶解的主要因素; 核燃料浸出实验结果显示铀和锕系镧系元素每天的浸出量是相应核素总量的1/107,比裂变产物的浸出速率小一个数量级。铁金属被各国选为高放废物处置容器材料的原因是其低价格、高强度和优秀的还原能力。在最不利的地下水侵入深地质处置库、近场处置容器防腐层破损的情景下,铁容器材料表面与地下水反应产生氢气,氢气通过还原反应消耗辐解产生的氧化性自由基和分子, 并能还原乏燃料表面的U（Ⅳ）,大幅度减缓乏燃料的腐蚀和溶解;乏燃料中裂变产物贵金属合金颗粒对氢气有催化作用;处置容器表面铁金属能还原沉积溶解的多价态核素U（Ⅵ）、Np（Ⅴ）、Tc（Ⅶ）、Se（Ⅳ）和Se（Ⅵ）。希望本文对我国确立以铁基金属为处置容器材料的包括乏燃料在内的高放废物深地质处置概念有参考作用。     

Towards a Swedish repository for spent fuel

Nuclear power is producing electricity for the benefit of society but is also leaving radioactive residues behind. It is our responsibility to handle these residues in a safe and proper manner. The development of a system for handling spent fuel from nuclear power plants has proceeded in steps. The same is true for the actual construction of facilities and will continue to be the case for the final repository for spent fuel and other types of long-lived wastes. The primary objective in constructing the repository will be to isolate and contain the radioactive waste. In case the isolation fails for some reason the multibarrier system should retain and retard the radionuclides that might come into contact with the groundwater. A repository is now planned to be built in two steps where the first step will include deposition of about 400 canisters with spent fuel. This first step should be finished in about 20 years from now and be followed by an extensive evaluation of the results from not only this particular step but also from the development of alternative routes before deciding on how to proceed. A special facility to encapsulate the spent fuel is also required. Such an encapsulation plant is proposed to be constructed as an extension of the existing interim storage CLAB. Finding a site for the repository is a critical issue in the implementation of any repository. The siting process started a few years ago and made some progress but is by no means yet completed. It will go on at least into the early part of the next decade. When the present nuclear power plants begin to be due for retirement there should also be some facilities in place to take permanent care of the long-lived radioactive residues. Progress in siting will be a prerequisite for success in our responsibility to make progress towards a safe permanent solution of the waste issue.     

Moving the Largest Capacity PWR Dual-Purpose Cask in the World from Gösgen NPP to the Zwilag Interim Storage Site

Abstract

The Swiss Gösgen nuclear power plant (NPP) has decided to use two different methods for the disposal of its spent fuel. (1) To reprocess some of its spent fuel in dedicated facilities. Some of the vitrified waste from the reprocessing plant will be shipped back to Switzerland using the new COGEMA Logistics, TN81 cask. (2) To ship the other part of its spent fuel to the central interim storage facility at Zwilag (Switzerland) using a COGEMA Logistics dual-purpose TN24G cask. The TN24G is the heaviest and largest dual-purpose cask manufactured so far by COGEMA Logistics in Europe. It is intended for the transport and storage of 37 pressurised water-reactor (PWR) spent fuel assemblies. Four casks were delivered by COGEMA Logistics to Gösgen NPP. Three transports of loaded TN24G casks between Gösgen and Zwilag were successfully pelformed at the beginning of 2002 using the new COGEMA Logistics Q76 wagon specifically designed to transport heavy casks. This article describes the procedure of operations and shipments for the first TN24G casks up to storage at Zwilag. The fourth shipment of loaded TN24G was due to take place in October 2002. The TN24G cask, as part of the TN24 cask family, proved to be a very efficient solution for Kemkraftwerk Gösgen spent fuel management.     

Status of the FRG activities on direct disposal of spent LWR-fuel

As stipulated by the German Atomic Energy Act, reprocessing is the reference waste management route for LWR's in the Federal Republic of Germany (FRG).Spent fuel disposal without reprocessing is being developed to technical maturity for those fuel elements for which reprocessing is either technically not feasible or economically not justifiable. The reference concept for direct disposal is the emplacement of large and heavily-shielded casks in drifts of a repository mine located in a salt dome. Moreover, a back-up solution is being pursued which results in smaller canisters which are emplaced in boreholes.The mining authorities have pointed out that the feasibility of direct disposal is to be demonstrated before a license for industrial scale deployment could be granted. Demonstration tests are necessary in the following areas: shaft transport of large and heavily shielded casks, handling of the casks in the repository and thermal and rock mechanics investigations with respect to the drift emplacement concept.The results of the demonstrations tests as well as the results from layout and optimization studies for a common repository for both reprocessing waste and spent fuel will be available early enough to be incorporated into the licensing procedure for the FRG's first repository for heat-generating nuclear wastes. This means that direct disposal of spent fuel not suitable for reprocessing could be introduced in the future in addition to the reprocessing and recycling waste management concept.     

核电厂放射性固体废物管理系统的研究与开发

目前我国在运核电厂和其他所有堆型（CPR1000、EPR和AP1000）的在建核电厂均缺少一套统一的放射性固体废物管理系统,缺乏对放射性固体废物从产生到最终处置的全周期跟踪管理。根据核电厂的放射性废物管理需求,研制了一套适合于各核电机型的核电厂放射性固体废物管理系统,对废物源项、处理、暂存、运输、处置全过程进行跟踪,使放射性废物管理安全、可控;研发了废物管理跟踪单和数据库,分析了废物管理工艺流程的逻辑关系,根据废树脂、浓缩液、废滤芯、检修废物等处理工艺分别设计了核素计算模型,可推算指定时刻的放射性水平,实现放射性废物数据的深度分析、应用以及对放射性废物安全管理的全过程追踪。研究成果已经在国内部分核电厂使用,有助于提高核电厂的放射性废物管理水平,具有较大的安全和社会意义。同时,该系统记录的数据有助于核电厂实现辐射防护优化设计和放射性废物最小化管理。