Storage and Disposal of Radioactive Wastes

(Updated March 2017)

  • Radioactive wastes are stored so as to avoid any chance of radiation exposure to people, or any pollution.
  • The radioactivity of the wastes decays with time, providing a strong incentive to store high-level wastes for about 50 years before disposal.
  • Storage of used fuel is normally under water for at least five years and then often in dry storage.
  • Disposal of low-level wastes is straightforward and can be undertaken safely almost anywhere.

Most low-level radioactive waste (LLW) is typically sent to land-based disposal immediately following its packaging for long-term management. This means that for the majority (~90% by volume) of all of the waste types, a satisfactory disposal means has been developed and is being implemented around the world.

Concentrating on intermediate-level waste (ILW) and high-level waste (HLW), many long-term waste management options have been investigated worldwide which seek to provide publicly acceptable, safe and environmentally sound solutions to the management of radioactive waste. Some countries are at the preliminary stages of their investigations whilst others such as Finland and Sweden have made good progress in their investigations to select publicly acceptable sites for the future disposal of waste. In Carlsbad, New Mexico in the USA, the Waste Isolation Pilot Plant (WIPP) disposal facility for defence-related transuranic wastes is in operation, underground in a salt formation.

The following table sets out the commonly accepted disposal options. When considering these, it should be noted that the suitability of an option or idea can be dependent on the wasteform, volume and radioactivity of the waste. As such, waste management options and ideas described in this section are not all applicable to different types of waste.

Commonly-accepted disposal options

Option Examples
Near-surface disposal at ground level, or in caverns below ground level (at depths of tens of metres)
  • Implemented for LLW in many countries, including Czech Republic, Finland, France, Japan, Netherlands, Spain, Sweden, UK and USA.
  • Implemented in Finland and Sweden for LLW and short-lived ILW.
Deep geological disposal
(at depths between 250m and 1000m for mined repositories, or 2000m to 5000m for boreholes)
  • Most countries with high-level and long-lived radioactive waste have investigated deep geological disposal and it is official policy in various countries (variations also include multinational facilities).
  • Implemented in USA for defence-related ILW.
  • Preferred sites for HLW/spent fuel selected in France, Sweden, Finland and USAa.
  • Geological repository site selection process commenced in UK and Canada.

Additional ideas have also been considered and discounted in the past (see section on Other ideas for disposal below).

Near-surface disposal 

The International Atomic Energy Agency (IAEA) definitionb of this option is the disposal of waste, with or without engineered barriers, in:

  • Near-surface disposal facilities at ground level. These facilities are on or below the surface where the protective covering is of the order of a few metres thick. Waste containers are placed in constructed vaults and when full the vaults are backfilled. Eventually they will be covered and capped with an impermeable membrane and topsoil. These facilities may incorporate some form of drainage and possibly a gas venting system.
  • Near-surface disposal facilities in caverns below ground level. Unlike near-surface disposal at ground level where the excavations are conducted from the surface, shallow disposal requires underground excavation of caverns but the facility is at a depth of several tens of metres below the Earth's surface and accessed through a drift.

The term near-surface disposal replaces the terms 'shallow land' and 'ground disposal', but these older terms are still sometimes used when referring to this option.

These facilities will be affected by long-term climate changes (such as glaciation) and this effect must be taken into account when considering safety as such changes could cause disruption of these facilities. This type of facility is therefore typically used for LLW and ILW with a radionuclide content of short half-life (up to about 30 years).

Near-surface disposal facilities are currently in operation in:

  • UK – Low Level Waste Repository at Drigg in Cumbria operated by UK Nuclear Waste Management Ltd (a consortium led by Washington Group International with Studsvik UK, Serco and Areva) on behalf of the Nuclear Decommissioning Authority.
  • Spain – El Cabril low and intermediate level radioactive waste disposal facility operated by ENRESA.
  • France – Centre de l'Aube operated by Andra.
  • Japan – Low-Level Radioactive Waste Disposal Center at Rokkasho-Mura operated by Japan Nuclear Fuel Limited.
  • USA – five low-level waste disposal facilities: Texas Compact facility near the New Mexico border, operated by Waste Control Specialists; Barnwell, South Carolina; Clive, Utah; Oak Ridge, Tennessee – all operated by EnergySolutions; and Richland, Washington – operated by American Ecology Corporation.

Near-surface disposal facilities in caverns below ground level are currently in operation in:

  • Sweden – the SFR final repository for short-lived radioactive waste at Forsmark, where the depth of the facility is 50m under the Baltic seabed – operated by the Swedish Nuclear Fuel and Waste Management Company (SKB).c
  • Finland – Olkiluoto and Loviisa power stations where the depth of the facilities are each at about 100 metres.d

Deep geological disposal

The long timescales over which some of the waste remains radioactive led to the idea of deep geological disposal in underground repositories in stable geological formations. Isolation is provided by a combination of engineered and natural barriers (rock, salt, clay) and no obligation to actively maintain the facility is passed on to future generations. This is often termed a multi-barrier concept, with the waste packaging, the engineered repository and the geology all providing barriers to prevent the radionuclides from reaching humans and the environment. In addition, deep groundwater is generally devoid of oxygen.

Mined repositories

The main concept is for a repository comprising mined tunnels or caverns into which packaged waste would be placed. In some cases (e.g. wet rock) the waste containers are then surrounded by a material such as cement or clay (usually bentonite) to provide another barrier (called buffer and/or backfill). The choice of waste container materials and design and buffer/backfill material varies depending on the type of waste to be contained and the nature of the host rock-type available.

Excavation of a deep underground repository using standard mining or civil engineering technology is limited to accessible locations (e.g. under land or nearshore), to rock units that are reasonably stable and without major groundwater flow, and to depths of between 250m and 1000m. At a depth greater than 1000m. The contents of the repository would be retrievable in the short term, and if desired, longer-term.

Deep geological disposal remains the preferred option for waste management of long-lived radioactive waste in several countries, including Argentina, Australia, Belgium, Czech Republic, Finland, Japan, Netherlands, Republic of Korea, Russia, Spain, Sweden, Switzerland and USA. Hence, there is much information available on different disposal concepts; a few examples are given here. The only purpose-built deep geological repository for long-lived ILW that is currently licensed for disposal operations is in the USA. Plans for disposal of spent fuel are well advanced in Finland, Sweden, France and the USA, though in the USA there has been a political setback. In Canada and the UK, deep disposal has been selected and the site selection process has commenced.

The Swedish proposed KBS-3 disposal concepte uses a copper container with a steel insert to contain the spent fuel. After placement in the repository about 500 metres deep in the bedrock, the container would be surrounded by a bentonite clay buffer to provide a very high level of containment of the radioactivity in the wastes over a very long time period. In June 2009, the Swedish Nuclear Fuel and Waste Management Company (SKB) announced its decision to locate the repository at Östhammar (Forsmark).

Finland's repository programme is also based on the KBS-3 concept. Spent nuclear fuel packed in copper canisters will be embedded in the Olkiluoto bedrock at a depth of around 400 metres. The country's nuclear waste management company, Posiva Oy, expects the repository to begin disposal operations in 2023. Its construction was licensed in November 2015.

The deposits of native (pure) copper in the world have proven that the copper also used in the final disposal container can remain unchanged inside the bedrock for extremely long periods, if the geochemical conditions are appropriate (reducing groundwaters). The findings of ancient copper tools, many thousands of years old, also demonstrate the long-term corrosion resistance of copper, making it a credible container material for long-term radioactive waste storage.

Deep boreholes

As well as mined repositories which have been the focus of international efforts so far, deep borehole disposal of high-level radioactive waste has been considered as an option for geological isolation for many years, including original evaluations by the US National Academy of Sciences in 1957 and more recent conceptual evaluations. In contrast to recent thinking on mined repositories, the contents would not be retrievable.

The concept consists of drilling a boreholes into crystalline basement rock to a depth of about 5000 metres, emplacing waste canisters containing used nuclear fuel or vitrified radioactive waste from reprocessing in the lower 2000 metres of the borehole, and sealing the upper 3000 metres of the borehole with materials such as bentonite, asphalt or concrete. The disposal zone of a single borehole could thus contain 400 steel canisters each 5 metres long and one-third to half a metre diameter. These might be emplaced in strings of 40 canisters. The waste containers would be separated from each other by a layer of bentonite or cement.

Boreholes can be readily drilled offshore (as described in the section below on sub seabed disposal) as well as onshore in host rocks both crystalline and sedimentary. This capability significantly expands the range of locations that can be considered for the disposal of radioactive waste.

Deep borehole concepts have been developed (but not implemented) in several countries, including Denmark, Sweden, Switzerland and USA for HLW and spent fuel. Compared with deep geological disposal in a mined underground repository, placement in deep boreholes is considered to be more expensive for large volumes of waste. This option was abandoned in countries such as Sweden, Finland and the USA. The borehole concept remains an attractive proposition for the disposal of smaller waste forms including sealed radioactive sources from medical and industrial applications.o

An October 2014 US Department of Energy report said: “Preliminary evaluations of deep borehole disposal indicate a high potential for robust isolation of the waste, and the concept could offer a pathway for earlier disposal of some wastes than might be possible in a mined repository.” In January 2016 the DOE commissioned a team led by Battelle to drill a 4880-metre test borehole into crystalline basement rock in North Dakota.

Disposal in clay, Europe

The Belgian disposal concept proposes that spent fuel and HLW is placed in high integrity steel containers and then emplaced in excavated tunnels 230 metres deep within a ductile (self-sealing) clayf – the Boom clay. The very low permeability of the clay leads to virtually no groundwater flow over long time periods. Waste would be backfilled with excavated clay or, alternatively, could be emplaced into unlined secondary tunnels where the clay would be allowed to creep into contact with the waste containers. Similar systems have been proposed in the Netherlands and, using less plastic clays, in France and Switzerlandg (Opalinus clay). Clay is generally suitable for heat-generating HLW, and the OECD Nuclear Energy Agency has a ‘Clay Club’ researching this.

The French radioactive waste disposal agency Andra is designing a deep geological repository in clays at Bure in eastern France. This will be for disposal of vitrified high-level waste (HLW) and long-lived intermediate-level waste. The repository is designed to operate at up to 90ºC, which is likely to be reached about 20 years after emplacement. Andra expects to apply for a construction and operating licence in 2014.

Yucca Mountain, USA

At the end of 1987, the Nuclear Waste Policy Act was amended to designate Yucca Mountain, located in the remote Nevada desert, as the sole US national repository for spent fuel and high-level waste from nuclear power and military defence programmes. An application by the US Department of Energy (DoE) to construct the repository was submitted in June 2008.

The repository would exist 300 metres underground in an unsaturated layer of welded volcanic tuff rock. Waste would be stored in highly corrosion-resistant double-shelled metal containers, with the outer layer made of a highly corrosion-resistant metal alloy, and a structurally strong inner layer of stainless steel. Since the geological formation is essentially dry, it would not be backfilled but left open to some air circulation. Drip shields made of corrosion-resistant titanium would cover the waste containers to divert possible future water percolation and provide protection from possible falling rock or debris. Containment relies on the extremely low water table, which lies approximately 300 metres below the repository, and the long-term durability of the engineered barriers.

The project has experienced many delays since its inception and following the 2009 presidential election the Barack Obama administration decided to cancel ithHowever, in June 2010, the Nuclear Regulatory Commission's Atomic Safety and Licensing Board (ASLB) rejected the DOE's motion to withdraw the licence application, and in August 2013 the federal Appeals Court ordered the NRC to resume its review of the DOE's application for a licence to construct and operate the Yucca Mountain repository. The final volumes of the NRC’s safety evaluation report were published early in 2015i.

Disposal in layered salt strata or domes

Geological salt environments have a very low rate (perhaps even absence) of groundwater flow and feature gradual self-sealing of the excavations due to creep of the salt, which is plastic. Salt is generally suitable for heat-generating HLW, and the OECD Nuclear Energy Agency has a ‘Salt Club’ researching this.

The Waste Isolation Pilot Plant (WIPP)j in New Mexico for defence transuranic wastes (long-lived ILW) has been operational since 1999. For this repository natural rock salt is excavated from a Permian layer several metres thick, between other types of rock, 650 metres below ground level. The wastes placed in these excavations contain large volumes of long-lived ILW, usually in steel drums. These are then placed on pallets and stowed in excavated rooms or caverns. The salt is plastic and will eventually seal the wastes and isolate them permanently. Containment of the radionuclides in the wasteform mostly relies on the almost complete absence of water flow in the salt. To July 2013, there had been 11,500 road shipments of wastes to WIPP from 12 DOE sites, and 87,500 cubic metres of ILW disposed2.

Salt environments are also available in northern Germany and the Netherlands although these are salt domes rather than bedded formations. In Germany, the former salt mines at Asse and Morsleben have been used for LLW and ILW disposal though this has now been suspended. The decommissioning process is now being investigated to determine the method for backfilling and sealing the repository.k

Following an exhaustive site selection process the state government of Lower Saxony in 1977 declared the salt dome at Gorleben to be the location for a German national centre for disposal of radioactive wastes. It is now considered a possible site for geological disposal of high-level waste. The site could be available as a final repository from 2025, with a decision to be made about 2019. Some €1.5 billion was spent over 1979 to 2000 researching the site. Work then stopped due to political edict, but resumption of excavation was approved following a change of government in 2009.l

Nirex Phased Disposal Concept, UK

The UK's Nirex Phased Disposal Concept (or Phased Geological Disposal Concept) has been developed for relatively large volumes of ILW and LLW, usually cemented into stainless steel containers.m These containers would be emplaced into a repository in a host rock environment below the water table. The waste would be monitored and remain retrievable and the groundwater managed to prevent contact with the wastes, until such a time that the repository is sealed. When this happens, the waste will be surrounded (backfilled) by specially formulated cement and the repository allowed to resaturate. The cement would provide a long lasting alkaline environment that contributes to containment of the waste by preventing many radionuclides from dissolving in the groundwater. Similar cement-based schemes for ILW disposal have been proposed in France, Japan, Sweden and Switzerland.

Multinational repositoriesn

Not all countries are adequately equipped to store or dispose of their own radioactive waste. Some countries are limited in area, or have unfavourable geology and therefore siting a repository and demonstrating its safety could be challenging. Some smaller countries may not have the resources to take the proper measures on their own to assure adequate safety and security, or they may not have enough radioactive waste to make construction and operation of their own repositories economically feasible.

It has been suggested that there could be multinational or regional repositories located in a willing host country that would accept waste from several countries. They could include, for example use by others of a national repository operating within a host country, or a fully international facility owned by a private company operated by a consortium of nations or even an international organisation. However, for the time being, many countries would not accept nuclear waste from other countries under their national laws. National policies towards radioactive waste management are listed in Waste Management in the Nuclear Fuel Cycle Appendix 3: National Policies and the information page on International nuclear waste disposal concepts.

Interim waste storage and transport

Specially designed interim surface or sub-surface storage waste facilities are currently used in many countries to ensure the safe storage of hazardous radioactive waste pending the availability of a long-term disposal option. Interim storage facilities are generally used for intermediate-level waste (ILW) and high-level waste (HLW), including used nuclear fuel from reactors.

Used fuel assemblies are shipped in Type B casks which are shielded with steel, or a combination of steel and lead, and can weigh up to 110 tonnes when empty. A transport cask may hold from 6 to 24 tonnes of used fuel. Some information on transport of used fuel is given below, because sometimes very similar equipment is used for storage, but more detail is in the information paper on Transport of Radioactive Material.

Storage ponds

Storage ponds at reactors, and those at centralised facilities such as CLAB in Sweden, are 7-12 metres deep, to allow several metres of water over the used fuel in racked fuel assemblies typically about 4 m long and standing on end. The multiple racks are made of metal with neutron absorbers incorporated in it. The circulating water both shields and cools the fuel. These pools are robust constructions made of thick reinforced concrete with steel liners. Ponds at reactors may be designed to hold all the used fuel for the life of the reactor, but usually the design assumes some removal of cooled fuel for reprocessing or to dry storage.

Some storage of fuel assemblies which have been cooling in ponds for at least five years is in dry casks, or vaults typically with air circulation inside concrete shielding. Transfer from wet storage to dry casks at a power plant site may use special shielded transfer casks, which are less robust than those used for transport beyond the site.

Multi-purpose canisters

Sealed stainless steel casks or multi-purpose canisters (MPCs), each holding up to 80 fuel assemblies with inert gas, are commonly used for transporting, storing and eventual disposal of the used fuel, inside robust overpacks – metal for transport, or mainly concrete for storage. Each MPC, of 13 mm welded stainless steel with a secure lid and internal fuel basket to hold and keep the fuel assemblies separate, is designed for up to 45 kW heat load. MPCs have standard external dimensions and the number of fuel assemblies actually loaded into one depends on their characteristics. Some are double-walled (DWC), with helium in between the layers. Once an MPC is loaded the contents should never need to be handled again.

Holtec’s MPC contains a 68-cell fuel basket for BWR fuel, a 24-cell flux-trap, or a 32-cell non-flux trap fuel basket for PWR fuel. Some have neutron-absorbing Metamic-HT fuel baskets and liners with high thermal conductivity, enabling relatively-hot three-year-old spent fuel to be placed in them.

Transport casks

For transport beyond the power plant site, the MPC is put inside a steel transport cask such as a version of Holtec HI-STAR (STAR = storage, transport and repository). This is classified as a Type B shipping cask, providing secure shielding and if necessary also some heat dissipation.

Holtec’s HI-STAR 100 is a high-capacity system which is engineered to accept one sealed multi-purpose canister (MPC) containing a 68-cell fuel basket for BWR fuel or a 32-cell fuel basket for PWR fuel. It can then transport its highly radioactive payload safely. The MPC containing the fuel can be transferred to HI-STORM 100 storage systems on an ISFSI pad, exchanging one versatile overpack for another which is engineered for maximum shielding. The MPC is loaded under water.

Areva has a range of Type B transport casks for used fuel, its newest ones being TN112 and TN117. Its older TN12/2 cask designed for used fuel has been adapted for fresh MOX fuel, and holds 12 PWR fuel assemblies or 32 BWR ones. It is robust, and with shock-absorbing covers at each end. Each year Areva moves about 200 casks with used fuel and about 20 with vitrified HLW. Fresh MOX fuel and vitrified HLW use different casks than for used fuel. Used MOX fuel requires greater neutron shielding, such as provided by the TN112 cask, which has a double containment barrier and can also take 50 kW heat load. It has lead shielding with steel, while the older one is simply forged steel. Both have resin compound shielding. They are both loaded under water.

Another Type B shipping package is Holtec’s HI-STAR 80 cask, a multi-layered steel cylinder which holds 12 PWR or 32 BWR high-burnup used fuel assemblies (above 45 GWd/t) which have had cooling times as short as 18 months. The HI-STAR 60 cask can transport 12 PWR used fuel assemblies. The HI-STAR 180 was the first one licensed to transport high-burnup fuel, and holds 32 or 37 (14x14) PWR used fuel assemblies. The HI-STAR 190 cask has 38 kW heat load capacity and one version is to be used domestically in Ukraine for PWR fuel. It is envisaged as the main used fuel transport to central storage or disposal sites in the USA. All HI-STAR casks have an aluminium impact limiter at each end, making the whole unit about 7 metres long.

Spent Fuel Canister

Holtec HI-STAR with MPC 37 for PWR fuel (24 tonnes)

In Russia, TUK (= transportation packaging set) casks are used to transport used nuclear fuel. Several TUK-13 casks fit into a container or TK carrier for rail transport, each cask holding about 6 tonnes of fuel. A TUK-1410 cask from the Federal Centre for Nuclear & Radiation Safety has now been licensed to replace the older model for VVER-1000 fuel, both in Russia and from overseas. Each weighs over 100 tonnes, holds 18 VVER fuel assemblies weighing 9 tonnes in a removable canister, and is designed for hotter fuel – up to 36 kW heat load. The TUK-14 series is also more heavily shielded, and can carry VVER-1200 fuel. It fits a TK-U-141 railway carrier.

OMZ subsidiary Izhorskiye Zavody has completed the prototype of a new generation of container for the transport and storage of used nuclear fuel from VVER-1000 and VVER-1200 reactors. The TUK-151 container is a thick-walled 2.5m diameter vessel with a sealed lid, a length of about 6 metres and a laden weight of 116 tonnes. Unlike the TUK-13 container design, the TUK-151 is not made of stainless steel, but has a low alloy steel-plated interior surface that offers "superior strength" and resistance to fracturing at low temperatures under dynamic loads.

In the UK 47- or 53-tonne rectangular Type B flasks have long been used to transport Magnox and AGR fuel, which is held in internal skips. In Sweden, more than 80 large transport casks are shipped annually to a central interim waste storage facility called CLAB. Each 80 tonne cask has steel walls 30 cm thick and holds 17 BWR or 7 PWR fuel assemblies. The used fuel is shipped to CLAB after it has been stored for about a year at the reactor, during which time heat and radioactivity diminish considerably. Both these systems are for transport only, not storage.

In the USA the Nuclear Regulatory Commission notes: “Over the last 40 years, thousands of shipments of commercially generated spent nuclear fuel have been made throughout the United States without causing any radiological releases to the environment or harm to the public.” Most of these shipments are between different power plants owned by the same utility, so as to share storage space for used fuel.

The internal structure of transport casks (using MPCs or not) is designed to maintain separation of fuel assemblies even in extreme accidents, and the external structure is designed to maintain safe containment in extreme accidents. Both features are tested before licensing.

Storage casks

For storage, each multi-purpose canister (MPC) is enclosed in a ventilated storage module or overpack made of concrete and steel. These are commonly standing on the surface, about 6 metres high, cooled by air convection, or they may be horizontal in banks, or they may be vertical and below grade, with just the tops showing. The modules are robust and provide full shielding. At a nuclear power plant, a shielded transfer cask is used to move the MPC from the used fuel pool to a storage module such as the vertical HI-STORM module (STORM = storage module). Holtec’s transfer casks are called HI-TRAC.

A free-standing above-ground system is Holtec’s HI-STORM 100, which accommodates a variety of sealed stainless steel MPCs vertically inside ventilated concrete and steel overpacks standing on a concrete pad. The overpack has 65 cm of concrete inside steel casing for shielding. This system is used at many US plants. A below-ground variant is HI-STORM 100U, and a more sophisticated version of this is UMAX (see below).

A larger canister HI-STORM FW storage module is a flood-and wind-resistant version with high seismic tolerance – 1.2 Gal. It holds 37 PWR assemblies, or 89 BWR ones, or 31 VVER-1000 ones, with maximum heat load 46 kW. Both kinds of canister can hold damaged fuel in special failed fuel containers.

A new large storage cask is HI-STORM MIC (Mega-Impact Capable) designed with EDF Energy in the UK and having a 100-year design life. It uses a double-walled MPC, and is heavily shielded.

HI-STORM CIS (Consolidated Interim Storage) is the next generation underground storage design that will house used fuel packaged in any canister supplied by any cask vendor. Each ventilated module will store two MPCs, vertically stacked to halve the required storage area. A six-hectare HI-STORM CIS ISFSI can store 4000 canisters containing more than 50,000 tonnes of used fuel.

Ukraine’s new Central Spent Fuel Storage Facility (CSFSF) will use ventilated Hi-Storm 190 casks for storage and Hi-Star 190 casks for transport to the site. Ukraine requires storage to be in double-wall canisters (DWC), and these are also used at Sizewell in the UK, giving an increased service life. In Ukraine, Holtec is establishing a centre for supply of dry storage and transport systems to the country's 15 operating reactors and possibly other VVERs around the world.

A sophisticated below-ground system is Holtec’s HI-STORM UMAX storage system, already deployed at two US nuclear power plant sites, and is proposed for a large consolidated ISFSI site in New Mexico for the Eddy Lea Energy Alliance. This stores the canisters containing used fuel in ventilated vertical steel and concrete cavity enclosure containers 5 metres high below ground, with massive lids at ground level. The containers are set up in a 7.6m deep excavation and low-strength concrete grout is backfilled around them. The final half metre of fill is a reinforced concrete pad. Seismic tolerance is about 2000 Gal. Holtec’s HI-STAR 190 casks will be used for transport in the USA, and will accommodate all types of spent fuel MPCs used in the USA. The UMAX is already operational at Calloway nuclear plant in the USA.

An above-ground horizontal system is Areva’s NUHOMS HSM-H system, used by 20 of the 67 ISFSIs at US nuclear plants. The used fuel is sealed in 15 mm thick stainless steel dry storage canisters which are then moved in horizontal shielded transfer casks to massive horizontal storage modules, with each hole 5 m long and 2 m diameter. Areva claims better heat distribution in these than vertical systems, using conduction more than convection, and also points to no gaps between modules being very safe seismically (1500 Gal) and radiologically. Each canister holds 32-37 PWR fuel assemblies or 61-69 BWR assemblies, in helium. Areva plans to have a US-licensed transport cask for NUHOMS by 2020, and meanwhile it has one for high-burnup fuel. NUHOMS horizontal storage can accommodate a variety of canister designs, and the Areva NUHOMS canisters are compatible with other, e.g. Holtec, vertical storage systems.

Germany’s Gesellschaft für Nuklear-Service mbH (GNS) set up in 1977 and owned by the four nuclear utilities there is both an operator of waste storage and supplier of two types of casks. The main type is CASTOR casks, more than 1000 of which are used in Germany alone for transport and interim storage of spent fuel and high-level wastes. The cask body provides full shielding and allows loading after very short cooling times of high-burnup fuel, and they are closed with two lids. GNS’s CONSTOR casks are similar but with concrete in the walls and desigend for cooler fuels.

Lithuania is using GNS CONSTOR M2 dry storage containers at Ignalina, each holding about 90 RBMK fuel assemblies. Bulgaria is also using 270 CONSTOR casks for VVER-440 fuel. CONSTOR casks are robust steel casks with concrete in the walls and three lids, cost-optimised for interim storage of spent fuel which has cooled. They are portable though evidently not designed for shipping used fuel. They are not ventilated and may have a concrete overpack.

Japan’s Federation of Electric Power Companies of Japan (FEPC) aims to develop an all-aluminum, thinner and lighter basket for a concrete cask that would be capable of storing more spent fuel assemblies at temperatures of up to 400°C. It wants to have the design ready for deployment by the mid-2020s. Most dry storage casks require fuel to be cooled in ponds until the temperature drops to about 250°C.

ISFSI, at site or central

A collection of casks or modules comprises an independent spent fuel storage installation (ISFSI), which in the USA is licensed separately from any associated power plant, and is for interim storage only. About one-quarter of US used fuel is stored thus.

In the USA a large ISFSI is proposed in Texas, using Areva’s NUHOMS system for up to 40,000 tonnes of used fuel. Waste Control Specialists applied for a licence for the facility in April 2016.

Zwilag’s ZZL in Switzerland and Ahaus and Gorleben in Germany are examples of operating long-term above-ground central interim dry storage for HLW.

Recognising that long-term management options, specifically for ILW and HLW, may require significant time to be achieved, interim storage arrangements may need to be extended beyond the time periods originally envisaged.

An indication of the scale of dry storage is that the market for dry storage casks in the Americas, where there is the highest demand, was over $600 million in 2015 and is expected to rise to over $1 billion by 2020.

Some countries, including Australia, Belgium, Netherlands, Germany, Italy and Switzerland also place low-level waste (LLW) in interim storage, although most LLW is typically sent directly to land-based near-surface disposal facilities (see section above on Near-surface disposal).

See also the section below on Long-term above ground storage.

Other ideas for disposal

Numerous options for long-term nuclear waste management have been considered in the past. The table below highlights a number of these.

Ideas Examples
Long-term above ground storage
  • Investigated in France, Netherlands, Switzerland, UK and USA.
  • Not currently planned to be implemented anywhere.
Disposal in outer space (proposed for wastes that are highly concentrated)
  • Investigated by USA.
  • Investigations now abandoned due to cost and potential risks of launch failure.
(proposed for wastes that are heat-generating)
  • Investigated by Russia, UK and USA.
  • Not implemented anywhere.
  • Laboratory studies performed in the UK.
Disposal at subduction zones
  • Investigated by USA.
  • Not implemented anywhere.
  • Not permitted by international agreements.
Sea disposal
  • Implemented by Belgium, France, Federal Republic of Germany, Italy, Japan, Netherlands, Russia, South Korea, Switzerland, UK and USA.
  • Not permitted by International agreements.
Sub seabed disposal
  • Investigated by Sweden and UK (and organisations such as the OECD Nuclear Energy Agency).
  • Not implemented anywhere.
  • Not permitted by international agreements.
Disposal in ice sheets (proposed for wastes that are heat-generating)
  • Investigated by USA.
  • Rejected by countries that have signed the Antarctic Treaty or committed to providing solutions within national boundaries.
Deep well injection
(for liquid wastes)
  • Implemented in Russia for many years for low- and intermediate-level wastes.
  • Investigations abandoned in the USA in favour of deep geological disposal of wastes in solid form.

Long-term above ground storage

Above ground storage is normally considered an interim measure for the management of radioactive waste (see section above on Interim waste storage). But it can be considered as effectively a disposal option. France investigated it for HLW within the framework of the 1991 law on research into radioactive waste management (Act No 91-1381 of 30 December 1991, also known as the 'Bataille Act' after the name of its proposer), but not as a means of final disposal. However, controlled surface storage over longer time periods (greater than a couple of hundred of years) has also been suggested as a long-term waste management option.

Long-term above ground storage involves specially constructed facilities at the Earth's surface that would be neither backfilled nor permanently sealed. Hence, this option would allow monitoring and retrieval at any time without excessive expenditure.

Suggestions for long-term above ground storage broadly fall into two categories:

  • Conventional stores of the type currently used for interim storage, which would require replacement and repackaging of waste every 200 years or so.
  • Permanent stores that would be expected to remain intact for tens of thousands of years. These structures are often referred to as 'Monolith' stores or 'Mausoleums'.

The latter category of store is derived from the principle of 'guardianship', where future generations continue to monitor and supervise the waste.

Both suggestions would require information to be passed on to future generations, leading to the question of whether the stability of future societies could be ensured to the extent necessary to continue the required monitoring and supervision.

No country is currently planning to implement long-term (i.e. greater than a few hundred years) above ground storage. However, France is investigating long-term interim storage, but not necessarily above ground.

Long-term above ground storage has been considered as part of the range of management concepts in Switzerland by EKRA (Expert Group on Disposal Concepts for Radioactive Waste). The expert group (EKRA) observed that it was unclear what additional steps would be necessary to show how the long-term above ground storage concept could be brought to the state of development comparable with that of geological disposal and they recommended geological disposal as the preferred option.

Disposal in outer space

The objective of this option is to remove the radioactive waste from the Earth, for all time, by ejecting it into outer space. The waste would be packaged so that it would be likely to remain intact under most conceivable accident scenarios. A rocket or space shuttle would be used to launch the packaged waste into space. There are several ultimate destinations for the waste which have been considered, including directing it into the Sun.

The high cost means that such a method of waste disposal could only be appropriate for separated high-level waste (HLW) – i.e. long-lived highly radioactive material that is relatively small in volume – rather than spent fuel. The question was investigated in the United States by NASA in the late 1970s and early 1980s. Because of the high cost of this option and the safety aspects associated with the risk of launch failure, it was abandoned.

Today only radioisotope thermal generators (TRGs) containing a few kilograms of Pu-238 are launched by NASA (see information page on Nuclear Reactors for Space).

Rock melting

The deep rock melting option involves the melting of wastes in the adjacent rock. The idea is to either produce a stable, solid mass that incorporates the waste or encases the waste in a diluted form (i.e. dispersed throughout a large volume of rock) that cannot easily be leached and transported back to the surface. This technique has been mainly suggested for heat generating wastes such as vitrified HLW (see Waste Management in the Nuclear Fuel Cycle – Appendix 1: Treatment and Conditioning of Nuclear Wastes) and host rocks with suitable characteristics to reduce heat loss.

The HLW in liquid or solid form could be placed in an excavated cavity or a deep borehole. The heat generated by the wastes would then accumulate resulting in temperatures great enough to melt the surrounding rock and dissolve the radionuclides in a growing sphere of molten material. As the rock cools it would crystallise and incorporate the radionuclides in the rock matrix, thus dispersing the waste throughout a larger volume of rock. There are some variations of this option in which the heat-generating waste would be placed in containers and the rock around the container melted. Alternatively, if insufficient heat is generated the waste would be immobilised in the rock matrix by conventional or nuclear explosion.

Rock melting has not been implemented anywhere for radioactive waste. There have been no practical demonstrations of the feasibility of this option, apart from laboratory studies of rock melting. In the late 1970s and early 1980s, the rock melting option at depth was taken forward to the engineering design stage. This design involved a shaft or borehole which led to an excavated cavity at a depth of 2.5 kilometres. It was estimated, but not demonstrated, that the waste would be immobilised in a volume of rock 1000 times larger than the original volume of waste.

Another early proposal was the design of weighted, heat-resistant containers of heat generating wastes such that they would continue to melt the underlying rock, and allow them to move downwards to greater depths with the molten rock solidifying above it. This alternative resembles similar self-burial methods proposed for disposal of HLW in ice sheets (see section below on Disposal in ice sheets).

In the 1990s, there was renewed interest in this option, particularly for the disposal of limited volumes of specialised HLW, particularly plutonium, in Russia and in the UK. A scheme was proposed in which the waste content of the container, the container composition and the placement layout would be designed to preserve the container and prevent the wastes becoming incorporated in the molten rock. The host rock would be only partially melted and the container would not move to greater depths.

Russian scientists have proposed that HLW, particularly excess plutonium, could be placed in a deep shaft and immobilised by nuclear explosion. However, the major disturbance to the rock mass and groundwater by the use of nuclear explosions, as well as arms control considerations, has led to the general rejection of this option.

Disposal at a subduction zone

Subduction zones are areas where one denser section of the Earth's crust is moving towards and underneath another lighter section. The movement of one section of the Earth's crust below another is marked by an offshore trench, and earthquakes occur adjacent to the inclined contact between the two plates. The edge of the overriding plate is crumpled and uplifted to form a mountain chain parallel to the trench. Deep sea sediments may be scraped off the descending slab and incorporated into the adjacent mountains. As the oceanic plate descends into the hot mantle, parts of it may begin to melt. The magma thus formed migrates upwards, some of it reaching the surface as lava erupting from volcanic vents. The idea for this option would be to dispose of wastes in the trench region such that they would be drawn deep into the Earth.

Although subduction zones are present at a number of locations across the Earth's surface they are geographically very restricted. Not every waste-producing country would be able to consider disposal to deep-sea trenches, unless international solutions were sought. However, this option has not been implemented anywhere and, as it is a form of sea disposal, it is therefore not permitted by international agreements.

Disposal at sea

Disposal at sea involves radioactive waste being shipped out to sea and dropped into the sea in packaging designed to either: implode at depth, resulting in direct release and dispersion of radioactive material into the sea; or sink to the seabed intact. Over time the physical containment of containers would fail, and remaining radionuclides would be dispersed and diluted in the sea. Further dilution would occur as the radionuclides migrated from the disposal site, carried by currents. The amount of radionuclides remaining in the sea water would be further reduced both by natural radioactive decay, and by the removal of radionuclides to seabed sediments by the process of sorption.

This method is not permitted by a number of international agreements.

The application of the sea disposal of LLW and ILW has evolved over time from being a disposal method that was actually implemented by a number of countries, to one that is now banned by international agreements. Countries that have at one time or another undertaken sea disposal using the above techniques include Belgium, France, Federal Republic of Germany, Italy, Netherlands, Sweden, Switzerland and the UK, as well as Japan, South Korea, and the USA. This option has not been implemented for HLW.

Sub seabed disposal

For the sub seabed disposal option radioactive waste containers would be buried in a suitable geological setting beneath the deep ocean floor. This option has been suggested for LLW, ILW and HLW. Variations of this option include:

  • A repository located beneath the seabed. The repository would be accessed from land, a small uninhabited island or from an offshore structure.
  • Burial of radioactive waste in deep ocean sediments.

Sub seabed disposal has not been implemented anywhere and is not permitted by international agreements.

The disposal of radioactive wastes in a repository constructed below the seabed has been considered by Sweden and the UK. In comparison to disposal in deep ocean sediments, if it were desirable the repository design concept could be developed so as to ensure that future retrieval of the waste remained possible. The monitoring of wastes in such a repository would also be less problematic than for other forms of sea disposal.

Burial of radioactive waste in deep ocean sediments could be made by two different techniques: penetrators or drilling placement. The burial depth of waste containers below the seabed can vary between the two methods. In the case of penetrators, waste containers could be placed about 50 metres into the sediments. Penetrators weighing a few tons would fall through the water, gaining enough momentum to embed themselves into the sediments. A key aspect of the disposal of waste to seabed sediments is that the waste is isolated from the seabed by a thickness of sediments. In 1986, some confidence in this process was obtained from experiments undertaken at a water depth of approximately 250 metres in the Mediterranean Sea. The experiments provided evidence that the entry paths created by penetrators were closed and filled with remoulded sediments of about the same density as the surrounding undisturbed sediments.

Wastes could also be placed using drilling equipment based on the techniques in use in the deep sea for about 30 years. By this method, stacks of packaged waste would be placed in holes drilled to a depth of 800 metres below the seabed, with the uppermost container about 300 metres below the seabed.

In the 1980s, the feasibility of the disposal of HLW in deep ocean sediments was investigated and reported by the Organisation for Economic Co-operation and Development. For this concept, radioactive waste would be packaged in corrosion-resistant containers or glass, which would be placed beneath at least 4000 metres of water in a stable deep seabed geology chosen both for its slow water flow and for its ability to retard the movement of radionuclides. Radionuclides that are transported through the geological media, to emerge at the bottom of the seawater volume, would then be subjected to the same processes of dilution, dispersion, diffusion and sorption that affect radioactive waste disposed of at sea (see section above on Disposal at sea). This method of disposal therefore provides additional containment of radionuclides when compared with the disposal of wastes directly to the seabed.

Disposal in ice sheets

For this option containers of heat-generating waste would be placed in stable ice sheets such as those found in Greenland and Antarctica. The containers would melt the surrounding ice and be drawn deep into the ice sheet, where the ice would refreeze above the wastes creating a thick barrier. Although disposal in ice sheets could be technically considered for all types of radioactive wastes, it has only been seriously investigated for HLW, where the heat generated by the wastes could be used to advantage to self-bury the wastes within the ice by melting.

The option of disposal in ice sheets has not been implemented anywhere. It has been rejected by countries that have signed the 1959 Antarctic Treaty or have committed to providing a solution to their radioactive waste management within their national boundaries. Since 1980 there has been no significant consideration of this option.

Direct injection

This approach involves the injection of liquid radioactive waste directly into a layer of rock deep underground that has been chosen because of its suitable characteristics to trap the waste (i.e. minimise any further movement following injection).

In order to achieve this there are two geological prerequisites. There must be a layer of rock (injection layer) with sufficient porosity to accommodate the waste and with sufficient permeability to allow easy injection (i.e. act like a sponge). Above and below the injection layer there must be impermeable layers that act as a natural seal. Additional benefits could be provided from geological features that limit horizontal or vertical migration. For example, injection into layers of rock containing natural brine groundwater. This is because the high density of brine (salt water) would reduce the potential for upward movement.

Direct injection could in principle be used on any type of radioactive waste provided that it could be transformed into a solution or slurry (very fine particles in water). Slurries containing a cement grout that would set as a solid when underground could also be used to help minimise movement of radioactive waste.

Direct injection has been implemented in Russia and the USA.

In 1957 extensive geological investigations started in Russia for suitable injection layers for radioactive waste. Three sites were found, all in sedimentary rocks. At Krasnoyarsk-26 and Tomsk-7 injection takes place into two porous sandstone beds capped by clays at depths up to 400 metres. Whereas at Dimitrovgrad injection has now stopped, but took place into a sandstone and limestone formations at a depth of 1400 metres. In total, some tens of millions of cubic metres of low-, intermediate- and high-level radioactive wastes have been injected in Russia.

In the USA, direct injection of about 7500 cubic metres of low-level radioactive wastes as cement slurries was undertaken during the 1970s at a depth of about 300 metres over a period of 10 years at the Oak Ridge National Laboratory, Tennessee. It was abandoned because of uncertainties over the migration of the grout in the surrounding fractured rocks (shales). In addition a scheme involving high-level waste injection into crystalline bedrock beneath the Savannah River Site in South Carolina was abandoned before it was implemented due to public concerns.


Radioactive material is produced as a waste product from the oil and gas industry and generally referred to as 'technologically enhanced naturally occurring radioactive materials' (Tenorm)p. In oil and gas production, radium-226, radium-228 and lead-210 are deposited as scale in pipes and equipment in many parts of the world. Published data3 show radionuclide concentrations in scales up to 300,000 Bq/kg for Pb-210, 250,000 Bq/kg for Ra-226 and 100,000 Bq/kg for Ra-228. However, scrap steel from gas plants may be recycled if it has less than 500,000 Bq/kg (0.5 MBq/kg) radioactivity (the exemption level)q. This level however is 1000 times higher than the clearance level for recycled material (both steel and concrete) from the nuclear industry, where anything above 500 Bq/kg may not be cleared from regulatory control for recycling.

The largest Tenorm waste stream is coal ash, with 280 million tonnes arising globally each year, and carrying uranium-238 and all its non-gaseous decay products, as well as thorium-232 and its progeny. This is usually just buried. However, the double standard means that the same radionuclide, at the same concentration, can either be sent to deep disposal (if from the nuclear industry) or released for use in building materials (if fly ash).

Further Information


a. In the USA, the Yucca Mountain site in Nevada has been chosen to site a deep geologic repository for disposal of high-level radioactive waste, but the project is beset by political interference. A licence application to construct the repository was submitted to the US Nuclear Regulatory Commission (NRC) by the US Department of Energy (DOE) on 3 June 2008. However, soon after entering office, the Barack Obama administration decided to cancel the project1. Later, in June 2010, the NRC's Atomic Safety and Licensing Board (ASLB) denied the DOE's motion to withdraw the licence application. The order by the ASLB noted the 1982 Nuclear Waste Policy Act (NWPA) "does not give the Secretary the discretion to substitute his policy for the one established by Congress in the NWPA." The ASLB concluded: "Unless Congress directs otherwise, DOE may not single-handedly derail the legislated decision-making process by withdrawing the Application. DOE’s motion must therefore be denied." [Back]

b. See the Near Surface Disposal page in the Waste Technology Section of the IAEA website ( for further information. [Back]

c. A brochure on SFR, the final repository for radioactive operational waste, is available from SKB. [Back]

d. Some information on the Finnish repositories for operating waste can be found in Nuclear Waste Management in Finland, Finnish Energy Industries (2007). [Back]

e. The Swedish repository programme is described in the SKB brochures Deep repository for spent nuclear fuel, SKB (2003) and Final repository for spent nuclear fuel (2008). More detailed technical information is in RD&D Programme 2007 - Programme for research, development and demonstration of methods for the management and disposal of nuclear waste, SKB Report TR-07-12 (2007). Information on the Finnish programme can be found in Nuclear Waste Management in Finland, Finnish Energy Industries (2007) and on the website of Posiva Oy ( [Back]

f. The SAFIR 2 report - which presented scientific and technical research on the possible final disposal of high-level and/or long-lived radioactive waste in deep clay layers. [Back]

g. The Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) has proposed three siting regions for the high-level waste repository (see Announcement of potential siting regions for deep geological repositories, Nagra 06/11/2008). See the Nagra website ( for information on management of nuclear waste in Switzerland; in particular Opalinus Clay Project: Demonstration of feasibility of disposal (“Entsorgungsnachweis”) for spent fuel, vitrified high-level waste and long-lived intermediate-level waste - Summary Overview, Nagra (December 2002). [Back]

h. The website of the Department of Energy's Office of Civilian Radioactive Waste Management ( states: "The President has made clear that Yucca Mountain is not an option for waste storage." However, it appears that the Yucca Mountain project cannot be cancelled without the approval of Congress (see also Note a above). [Back]

i. See Note a above. [Back]

j. The website of the Waste Isolation Pilot Plant is at [Back]

k. The website of the German Federal Office for Radiation Protection (Bundesamt für Strahlenschutz, BfS - see contains information on Morsleben. For Asse, see and [Back]

l. The website of the German Federal Office for Radiation Protection (Bundesamt für Strahlenschutz, BfS - see contains information on Gorleben. [Back]

m. Nirex was incorporated into the Radioactive Waste Management Directorate of the UK's Nuclear Decommissioning Authority (NDA) in 2007 and no longer exists as a separate entity. The Nirex Phased Disposal Concept is outlined in the Introductory Leaflet What is the Nirex Phased Disposal Concept?, Nirex (2002). The NDA's strategy of research and development linked to the implementation of a geological disposal facility for higher activity radioactive wastes is given in The NDA’s Research and Development Strategy to Underpin Geological Disposal of the United Kingdom’s Higher-activity Radioactive Wastes, NDA Report NDA/RWMD/011 (March 2009). [Back]

n. For more information on multinational repositories, see the International Nuclear Waste Disposal Concepts information page and the Association for Regional and International Underground Storage (Arius) website ( [Back]

o. See Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel (October 2014), and  Disposal options for disused radioactive sources, International Atomic Energy Agency, Technical reports series, STI/DOC/010/436 (July 2005; ISBN: 9201003056). [Back]

p. In the UK, much of these wastes are exempt from the need for their disposal to be authorised under the UK's Radioactive Substances Act 1993 because of their low levels of radioactivity. However, some of the wastes are of higher activity and there are currently a limited number of disposal routes available. This includes re-injection back into the borehole (i.e. well-head), which is authorised by the UK's Environment Agency. [Back]

q. The main radionuclide in scrap from the oil and gas industry is radium-226, with a half-life of 1600 years as it decays to radon. [Back]


1. Obama dumps Yucca Mountain, World Nuclear News (27 February 2009); Yucca Mountain 'terminated', World Nuclear News (8 May 2009). [Back]

2. WIPP web site [Back]

3. Malcolm B. Cooper, Naturally Occurring Radioactive Materials (NORM) in Australian Industries - Review of Current Inventories and Future Generation, ERS-006, A Report prepared for the Radiation Health and Safety Advisory Council (Revision of September 2005) [Back]

General sources

Geological Waste Disposal page in the Waste Technology Section of the IAEA website.

Fact Sheet Understanding the potential for volcanoes at Yucca Mountain, Office of Civilian Radioactive Waste Management, US Department of Energy

Deep Borehole Disposal of Nuclear Waste: Final Report, Sandia Report SAND2012-7789, Sept 2012

Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel, October 2014, US DOE

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