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.     

Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 °C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.     

The Swiss Concept—Container Design and the Transport of L/ILW to the Planned Final Repository

Abstract

The design of the Swiss final repository for short lived L/ILW is based on a Nagra container and package concept. The package handling operations have been restricted to a minimum through the design of special handling tools. e.g. a gripper for 9 drums. The routine transport weight by rail is 56 t, and for non-routine transport 80 t (maximum). The transport of drums and reprocessing waste will be in re-usable steel containers and that of decommissioning waste in dual purpose transport and disposal containers. Most of the containers have standardised dimensions and corner fittings which are based on the ISO dimensions. The modes of transport for the containers and packages within the repository include overhead cranes, an air cushion platform for precise manoeuvering in limited spaces and internal rail transport. The handling and transport will mostly be remotely controlled and monitored by video cameras from the control room. Hence, the exposure times of the operating personnel in the radiation environment is minimised.     

Cost efficiency analysis of disposal alternatives in Korea

This research presents the results of calculating the disposal cost efficiency for the four disposal alternatives for the CANDU spent fuel that are under development in Korea currently. The KRS-1 alternative, developed first, was set as the standard, and the efficiency of the KRS-1 alternative was assumed to be 100%.The cost calculation result shows that the A-KRS-22, which was developed most recently among the CANDU spent fuel disposal alternatives, manifested −61.7%, −45.7%, −47.0%, −78.9% and −61.7% when compared to the KRS-1 alternative concerning disposal tunnel excavation, disposal hole excavation, bentonite, disposal canister and backfilling.Moreover, the cost calculation method for the dominant cost driver that uses the unit disposal module concept for the calculation of cost efficiency was used. As for the reason that the standard for efficiency measurement was taken per each bundle, it is because the amount of bundle capacity concerning the spent fuel differs by disposal canister.     

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.     

Criticality calculations of spent fuel in deep geological repository

Criticality calculations have been performed for a typical spent fuel disposal canister model filled with PWR fuel elements. Geometric and material properties of the disposal canister and disposal holes were modeled based on the Swedish preliminary disposal concept. Direct disposal of 5% enriched 16 × 16 PWR fuel was considered. We performed the calculations of the neutron multiplication factor using various disposal configurations, depending on the initial enrichment, fuel burnup, canister geometry and disposal holes configuration. The results showed that under normal conditions, when the canister is filled with fresh spent nuclear fuel, the system is deeply sub-critical. If it is assumed that the canister is faulty, leaking and filled with ground water, the system may become critical in the case of fresh fuel.     

Characteristics of a Geological Disposal System for the Increasing Burn-up of Spent Nuclear Fuel in Korea

The characteristics of a geological disposal system that can accommodate increasingly higher burn-up levels of spent fuel were assessed based on the Korea reference disposal system concept. First, a status investigation that included a projection of spent fuel quantity versus burn-up was carried out to demonstrate the trend toward higher burn-up levels. Next, the main features of the Korea reference disposal system were introduced. Finally, the disposal tunnel length, excavation volume, and raw materials (e.g., a cast insert, copper, bentonite and backfill) necessary for a disposal system were comprehensively analyzed to define the characteristics and overall effects on geological disposal at increasingly higher burn-up levels. Our study determined that it is reasonable to use a canister containing 4 spent fuel assemblies with burn-up levels up to 50GWD/MTU, while a canister containing 3 spent fuel assemblies can accommodate burn-up levels beyond 50GWD/MTU. A remarkable increase of 33% in disposal tunnel length and that of 30% in excavation volume were observed as the burn-up increased from 50 to 60GWD/MTU. However, this was offset by a reduction of 17% in raw materials used in canister fabrication. Therefore, it seems that spent fuel at increasingly higher burn-up levels is not a serious concern for deep geological disposal in Korea.     

Transporting radioactive materials to a geological disposal facility: an integrated approach

The Radioactive Waste Management Directorate (RWMD) of the UK’s Nuclear Decommissioning Authority (NDA) has responsibility for planning and implementing a Geological Disposal Facility (GDF) in the UK. The responsibility for demonstrating and providing a safe transport operation will be shared between a number of organisations acting as consignors, carriers and the consignee (the GDF operator). The radioactive waste transport system is national in scope. Its main objective is to deliver packaged waste to a facility for disposal in a manner that is safe, secure, planned, timely, cost effective, flexible, environmentally sound and robust against future changes. To fully appreciate the implications for ensuring transport safety a better understanding of the range of options for a GDF transport system is required. One extreme, the current planning assumption in the UK is that each waste producer (consignor) is individually responsible for organising their own transport to a GDF. The other extreme is where a single organisation is responsible for the provision of the transport system (an integrated transport service). Intermediate options will exist where the actual implementation could be anywhere on the scale between the two extremes. A fundamental issue for a GDF transport system for the delivery of Intermediate Level Waste (ILW), High Level Waste (HLW) and spent fuel is the timescale between initial waste packing and final sentencing to the repository. ILW, HLW and spent fuel will need to be managed until a GDF is available and delivery is confirmed. The timescales could be over 65 years given current assumptions. This paper reviews the feasibility of an integrated transport service for the delivery of ILW, HLW and spent fuel to a GDF. It defines the key elements of the integrated transport service, highlighting the advantages and disadvantages. Finally, it sets out the key considerations to be addressed during packing of wastes which will not be transported for up to 65 years.     

交联聚乙烯HIC在美国处置场应用实践及对我国的启示

简要介绍了从美国引进的交联聚乙烯高完整性容器(HIC)及其在美国Clive和Barnwell处置场的应用实践,对交联聚乙烯HIC在我国应用中存在的自由水含量、累积剂量、辐照分解气体、外包装等主要问题和解决思路进行了探讨,期望能对交联聚乙烯HIC在我国的处置方案的选择提供一些参考。     

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.     

Spent fuel transportation risk assessment: cask impact analyses

Abstract

The Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered the response of three certified casks to a range of impact accidents in order to determine whether or not they would lose their ability to contain the spent fuel or maintain effective shielding. The casks consisted of a lead shielded rail cask that can be transported either with or without an inner welded canister, an all-steel rail cask that is transported with an inner welded canister, and a DU shielded truck cask that is transported with directly loaded fuel. Finite element analyses were performed for impacts at speeds of 48, 97, 145 and 193 kilometres per hour into a rigid target. Impacts in end-on, side-on, and CG-over-corner orientations were analysed for each cask and impact speed. Calculations were performed to equate these impacts onto rigid targets with higher speed impacts onto the yielding targets that exist in the real world. These analyses indicated that a cask with an inner welded canister or a truck cask would not release radioactive material in any impact accident and that only very high-speed impacts onto hard rock targets could result in either release of material or significant degradation of shielding for rail casks without an inner canister. Impacts other than those onto flat unyielding targets were also considered. Analyses show that an impact that bypasses the impact limiters on the ends of the casks does not result in seal failure and neither does an impact by a locomotive also between the impact limiters.     

乏燃料密封容器开盖及其内容物回取技术研究

对于采用干湿法贮存的乏燃料而言,其后处理时面临的最大问题是如何安全高效地将乏燃料等内容物从封焊的密封容器中取出。针对这一问题,基于乏燃料密封容器及其内容物的结构特点,开展了乏燃料密封容器开盖及内容物回取技术研究,综合考虑切割热室使用环境、内容物回取后的收集和转移以及产生废物的收集处理等因素,制定了合理可行的开盖及回取工艺,研制了用于开盖和筒体分段切割的解体装置以及回取和吊装工具,并通过试验验证了工艺的可行性以及研制的工装具的可用性。      

Experience in the European Transport of Spent Nuclear Fuel

Abstract

Nuclear Transport Limited has been responsible for most of the European transport of spent nuclear fuel which has taken place to date, and therefore has unique experience in the field. The services and experience of Nuclear Transport Limited cover a whole range of flask types, large quantities of fuel transported, the design and provision of handling equipment, the logistics of operation, the arrangements necessary to maintain high standards of safety, and the need to alleviate public concern. Transport routes and communications systems have been developed, crossing national boundaries in Europe using road, rail and sea, currently employing the freight rail ferry for the short haul crossing of the English Channel, in contrast to the special purpose ships which are operated by PNTL for the long haul crossing of the ocean from Japan to the United Kingdom. Whilst there has never been a serious accident involving a spent fuel flask, procedures have been established to ensure an effective response to any accident or incident. This approach is consistent with the standards of safety that apply throughout the nuclear industry.     

Early age behaviour of concrete supercontainers for radioactive waste disposal

Various types of radioactive waste were and are produced in Belgium. This waste originates from different producers: nuclear power plants, medical applications, industry, research centre, etc. During the past 25 years several preliminary repository designs were proposed. Today, the cylindrical supercontainer is considered to be the most promising Belgian design on the matter of enclosing the vitrified high level radioactive waste (HLW) and the spent fuel assemblies and is based on the use of an integrated waste package composed of a carbon steel overpack surrounded by an Ordinary Portland Cement buffer. For the choice of this cementious buffer two compositions, a self-compacting concrete (SCC) and a traditional vibrated concrete (TVC), are being considered, tested and compared by means of an intensive laboratory characterization program. Through-going cracks in the concrete buffer should, at all times, be avoided because they will considerably ease the transport mechanisms inside the supercontainer. Therefore, finite element simulations are performed, using a 2.5-D thermal and crack modelling program, to predict the mechanical and thermal behaviour of the concrete buffer at any time during hardening. Looking at the finite element simulation results of the first stage of manufacturing of the supercontainer (cast in one), and the emplacement of the heat-emitting waste canister (second stage), we experience no early age cracking of the concrete buffer. The impact of environmental conditions and shrinkage and creep behaviour on the simulation results are noticeable.     

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.     

一种干式外圆机械切割装置的研制

在乏燃料后处理中，需要回取已封装在乏燃料贮存容器中的乏燃料。根据热室使用环境及乏燃料贮存容器的特点，从耐辐射设计、乏燃料贮存容器固定、切割进给、切割刀具及刀具更换、放射性废物最少化等方面进行设计响应，研制了一种在热室内开启乏燃料贮存容器的干式外圆机械切割装置。功能性试验验证了该装置满足设计和使用要求。      

The Receipt and Processing of Spent Fuel Transport Flasks at Sellafield

Abstract

British Nuclear Fuels plc (BNFL) has 40 years' experience of transporting spent fuel from UK and overseas customers to the Sellafield Site in Cumbria, northwest England. Fuel from reactor stations has been received at the site's pond storage facilities in a range of flask designs, from the cuboid flask designs used for Magnox and advanced gas-cooled reactor (AGR) fuels to the Excellox cylindrical flasks containing sealed multi-element bottles (MEBs). Light water reactor (LWR) fuels were first received at Sellafield in the early 1970s. BNFL have used numerous designs of flasks to successfully deliver thousands of tonnes of LWR spent fuel from overseas via the transport operating companies Nuclear Transport Limited (NTL), for Europe, and Pacific Nuclear Transport Limited (PNTL) for Japan. Similar flasks have been used for the internal transport of fuel between the Sellafield storage ponds. The associated receipt and despatch facilities therefore provide the key interface between the transport and pond storage modes prior to reprocessing the fuel. This paper focuses on operations associated with the receipt and processing of the transport flasks at Sellafield, demonstrating that the procedures used over several decades ensure that standards of safety and quality are maintained. It covers the operations associated with processing flasks and the maintenance regimes employed to ensure that the returned package meets IAEA transport regulations. In particular it focuses on contamination control, inspection regimes and the Sellafield plant safety case requirements.     

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.     

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.     

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.