pursuing opportunities for waste minimisation re-use and recycling waste treatment packaging storage transport and then final disposal where required.Sources of the UKs radioactive wastes including from power production, military programmes, medical uses and research reactors are described along with options for managing controlled wastes from pretreatment, treatment, conditioning and storage stages through to transportation to final disposal. Immobilisation (wasteform), temporary storage and permanent disposal options including near surface, deep and very deep geological disposal are covered.
Radioactive waste is material that contains, or is contaminated with radionuclides at concentrations or activities greater than the clearance levels set by the regulators, and for which no use is foreseen. The hazard associated with radioactive wastes depends on the concentration and nature of the radionuclides with those emitting higher energy radiation or being more toxic to life, being the most hazardous. Radioactive waste is accompanied by significant levels of radiation hence it requires not only immobilisation to prevent radionuclides spreading around the biosphere, but also shielding and, in some cases, remote handling. A waste with activity concentrations equal to, or less than, clearance levels is considered non-radioactive. Radioactive wastes are either controlled or uncontrolled (Lee et al. 2013a).
Controlled wastes are largely a product of the nuclear fuel cycle (NFC) used to generate electricity for civil use. Wastes are generated during ore mining and processing to access the uranium metal or oxide, its enrichment and synthesis into fuel (the front end of the nuclear fuel cycle), the operation and running of the reactor (operations wastes) and from fuel removal, treatment and disposal (the back end of the fuel cycle). Front end waste is contaminated basically with naturally occurring radionuclides whereas operational waste also contains fission and activated products (typically Low Level Waste, LLW and to a lesser extent Intermediate Level Waste, ILW). Front end wastes include contaminated mining wastes and uranium hexafluoride tails from enrichment. Operations wastes include spent filters and ion exchange resins, evaporator concentrates and absorber rods. Back end wastes include sludges from storage ponds, typically cemented ILW and vitrified HLW from reprocessing or spent fuel if direct disposal is planned. During the early part of the nuclear era consideration was not given to disposal of radioactive waste. As a result some NFC wastes (now termed legacy or historic wastes) are ill-characterised and stored under conditions which are far from ideal. They comprise a vast range of materials e.g. Pu-contaminated materials (PCM) such as paper, wood and plastics, fuel cladding, damaged and corroded fuel elements, old tools and equipment and assorted test samples often mixed together. Sometimes these have been stored under water and have degraded over time to form complex sludges and supernatant liquids. Controlled non-NFC wastes include those from various applications of radionuclides in research, medicine and industry including spent sealed radioactive sources (SRS) of isotopes used in medical applications.
Uncontrolled wastes arise when unexpected events occur or where the level of care was not taken that would be expected today. At Sellafield in the UK some materials were stored in inappropriate open ponds where ingress of atmospheric (salty) rain and organic matter (bird droppings etc.) has added to the complexity of the problem. Uncontrolled wastes also arise from accidents such as at Chernobyl, Ukraine and Fukushima, Japan. The financial cost of cleaning up such sites and others where accidental releases of radioactivity have occurred such as in Fukushima is enormous. Nonetheless, the nuclear industry is now developing smart clean-up programmes and concepts and the knowledge gained from these mistakes has helped us be more proactive in dealing with uncontrolled waste.
For practical purposes radioactive waste is classified in different classes depending on actual management needs. The IAEA classification scheme defines five classes of radioactive waste (IAEA, 2009): Very Short Lived Waste (VSLW), Very Low Level Waste (VLLW), Low Level Waste (LLW), Intermediate Level Waste (ILW) and High Level Waste (HLW).
VSLW is that radioactive waste which can be stored for decay over a limited period of no longer than a few years with subsequent clearance from regulatory control. Clearance is carried out according to existing national arrangements, after which VSLW can be disposed of, discharged or used. VSLW includes waste containing primarily radionuclides with very short half-lives which are most often used for research and medicine.
VLLW is that radioactive waste which does not necessarily meet the criteria of Exempt Waste (EW), but that does not need a high level of containment and isolation and so is suitable for disposal in near surface landfill type facilities with limited regulatory control. Typical VLLW includes soil and rubble with low levels of activity concentration.
LLW has higher activity contents compared to VLLW but with limited amounts of long lived radionuclides in it. Such waste requires robust isolation and containment for periods of up to a few hundred years and is suitable for disposal in engineered near surface facilities. LLW covers a very broad range of waste with long lived radionuclides only at relatively low levels of activity concentration.
ILW is that radioactive waste that, because of its radionuclide content, particularly of long lived radionuclides, requires a greater degree of containment and isolation than that provided by near surface disposal. However, ILW needs no provision, or only limited provision, for heat dissipation during its storage and disposal. ILW may contain long lived radionuclides, in particular, alpha emitting radionuclides that will not decay to a level of activity concentration acceptable for near surface disposal during the time for which institutional controls can be relied upon. Therefore ILW requires disposal at greater depths, of the order of tens of metres to a few hundred metres. Waste acceptance criteria for a particular near surface disposal facility depend on its actual design and operation (e.g. engineered barriers, duration of institutional control, site specific factors). A limit of 400 Bq/g on average and up to 4000 Bq/g for individual packages for long lived alpha emitting radionuclides has been adopted in many countries. For long lived beta and/or gamma emitting radionuclides, such as 14C, 36Cl, 63Ni, 93Zr, 94Nb, 99Tc and 129I, the allowable average activity concentrations may be considerably higher (up to tens of kBq/g) although they are specific to the site and disposal facility (IAEA, 2009).
HLW is the radioactive waste with levels of activity concentration high enough to require shielding in handling operations and generate significant quantities of heat by the radioactive decay process typically above several W/m3. HLW can be also the waste with large amounts of long lived radionuclides that need to be considered in the design of a disposal facility. Disposal in deep, stable geological formations usually several hundred metres or more below the surface is the generally recognized HLW disposal option. The IAEA classification scheme is rather generic and has no exact limits in defining radioactive waste classes. In the UK radioactive wastes are classified as VLLW, LLW, ILW and HLW (Table 1, DEFRA, 2008).
Table 1. Classification of radioactive waste in the UK.
| Class | Description |
| VLLW, Low volume | Wastes which can be disposed of with ordinary refuse, each 0.1 m3 of material containing less than 400 kBq of beta/gamma activity and is mostly comprised of small volumes from hospitals and universities. For carbon-14 and tritium containing wastes, the activity limit is 4000 kBq for each 0.1 m3 in total. |
| VLLW, High volume | Radioactive waste with an upper limit of 4 MBq per tonne (not including tritium) that can be disposed to specified landfill sites. For tritium containing wastes, the upper limit is 40 MBq per tonne. |
| LLW | Containing radioactive materials other than those suitable for disposal with ordinary refuse, but not exceeding 4 GBq per tonne of alpha or 12 GBq per tonne of beta/gamma activity. |
| ILW | Wastes with radioactivity levels exceeding the upper boundaries for LLW, but which do not need heating to be taken into account in the design of storage or disposal facilities. |
| HLW | Wastes in which the temperature may rise significantly as a result of their radioactivity, so this factor has to be taken into account in designing storage or disposal facilities. |
Waste generated during the operation of a Nuclear Power Plant (NPP) is generated mainly by treatment of water from the reactor or ancillaries including SF storage ponds and some decontamination operations. Standard effluent treatment technologies are based on evaporation (distillation), ion exchange, filtration or centrifuging. Typical process wastes from pressurized water reactors (PWR) are borated water concentrates, sludge or filter cartridges, and organic bead resin ion-exchangers (blow-down resins) from primary and secondary circuits whereas those from boiling water reactors (BWR) are water concentrates and sludge containing different types of ion exchange or filter media as organic powdered resins, diatomaceous earth, activated carbon, cellulose and organic bead resins. Maintenance waste is mainly solid comprising spent or damaged and contaminated equipment which cannot be repaired or recycled, and items such as contaminated clothes from operators, cardboard, bags, tools and plastic sheeting from maintenance work. Maintenance waste arises also from dismantling the internal structures of the reactor core including the used control rods. Liquid technological wastes comprise mainly oils and small amounts of lubricants and organic solvents used for decontamination. Typically the main radioactive contaminants in operational waste are short-lived radionuclides such as 60Co, 90Sr, 134Cs and 137Cs although long-lived radionuclides can be present in the internal elements of reactors.
Reactor fuel is typically in the form of ceramic Pu/U oxide pellets in a zircaloy or stainless steel metal rod. When the fuel reaches the end of its useful life, it is removed from the reactor and is considered as Spent Fuel (SF). SF typically contains about 95% 238U, about 3% of fission products and transuranic isotopes, about 1% Pu and 1% 235U. In the Open NFC the SF is considered as waste and can itself serve as a final wasteform since it is a reasonably stable solid providing it is encapsulated in an additional immobilising barrier such as a corrosion-resistant copper or lead container. The ceramic UO2 matrix of nuclear fuel retains the radionuclides and non-volatile fission products in its open fluorite crystal structure and its polycrystalline microstructure. The metal cladding of the fuel also, if intact, provides an additional barrier. About 30 tonnes of SF-waste are typically produced per year by a typical 1GW NPP.
In the Closed NFC SF is reprocessed to extract useful uranium and plutonium. Several reprocessing facilities currently are in operation worldwide including those at Sellafield (UK), La Hague (France) and Chelyabinsk (Russia). These were initially set up to extract material for weapons programmes but the Pu and U can be recycled for use in PWRs as mixed oxide (MOX) fuel. Reprocessing of SF involves removal of the fuel rod metal cladding followed by dissolution of the remainder in nitric acid, followed by chemical solvent extraction of the uranium and plutonium formed during the fuel burn-up via the Purex (Plutonium and Uranium Extraction) process. The remnant solution is HLW and contains the dissolved fission products (FPs) together with impurities from the cladding materials, inactive process chemicals, transuranic elements formed by neutron capture, and traces of un-separated plutonium. HLW is concentrated by evaporation to reduce the volume and stored in aqueous nitric acid solution in stainless steel tanks. HLW’s contain a host of products ranging from uranium fission products to fuel alloying elements including F, Al, Si and Mo; cladding elements including Zr, Mo, Nb, and Mg; transuranic elements including Np, Am, Cm and residual Pu. HLW’s also contain some of the process chemicals including kerosene, tributyl phosphate and related organic materials.
Decommissioning wastes are generated at the end of operation of NFC facilities including nuclear reactors. As well as waste from the radioactive ceramic fuel some structural materials become activated by elements undergoing neutron capture. The high alloy steel end caps from each fuel bundle in an Advanced Gas-cooled Reactor (AGR), for example, become so activated they are treated as HLW; since there are ~100 of these in each AGR assembly and a complete refuel occurs every two years the amount of waste is significant. The alloying elements of particular concern in steels are Co, Nb, Ni and Mo. After the SF is removed the NFC facilities must be decommissioned, demolished and eventually returned to green field or brown field use. During this process, large volumes of wastes are generated, although most is not radioactive. The amount of waste arising from decommissioning a typical NPP is 10,000 −15,000 tonnes. Much of this waste is concrete and other building material containing only small amounts of radioactivity. About a tenth of the decommissioning waste contains some radioactivity up to the intermediate level.
Many institutions worldwide installed small reactors in support of research and development programmes. Research reactors provide a wide range of training, research, commercial and nuclear power programme support functions from nuclear reactors which are generally not used for power generation. Their output (neutron beams) is used for non-destructive testing, analysis and testing of materials, production of radioisotopes, research and public outreach and education. The UK has had more than 30 research reactors since the 1950s but like all countries it has shut and decommissioned almost all of them. Its last remaining one CONSORT, owned and operated by Imperial College London, shut down in December 2012. Many of these reactors used novel fuels and decommissioning them requires programmes of research to determine suitable ways of managing their wastes. The Nuclear Decommissioning Authority (NDA) in the UK has a special programme examining options for these so-called exotic fuels (NDA 2012). Credible options for management of these fuels in the near term have been identified but significant R&D may be needed to identify routes to permanent disposal (NDA 2012).
Types and volumes of waste from applications of radionuclides in research, medicine and industry vary extensively in radiochemical, chemical and physical content. Research establishments are often involved in monitoring the metabolic or environmental pathways associated with materials as diverse as drugs, pesticides, fertilizers and minerals. The radionuclides most commonly employed in studying the toxicology of many chemical compounds and their associated metabolic pathways are 14C and 3H, as they can be incorporated into complex molecules with considerable uniformity. 125I has proved valuable in protein labelling. A spectrum of other radionuclides is available for research.
Most of the radioactive waste generated by nuclear research centres contains mainly short-lived radionuclides although long-lived radionuclides such as 14C, fissile radionuclides and transuranic elements may also be present. The main applications of radionuclides in medicine are in radio-immunoassays, radiopharmaceuticals, diagnostic procedures and radiotherapy. The radionuclides used in hospitals for medical diagnostic procedures and treatments are very short lived, and the waste generated is usually stored for decay before further treatment as non-radioactive waste. Positron Emission Tomography (PET), for example, incorporates cyclotron-generated 11C (20 minute half-life) or 18F (110 minute half-life) in a molecule such as sugar which is intravenously administered to the patient and is detected during its circulation around the body. Some radionuclides used in medical applications however have longer half-lives including 57Co (271.7 days) used in clinical measurements and 3H (12.3 years) and 14C (5730 years) used in radio-labelling. Medical applications of radionuclides such as for bone densitometry, manual brachytherapy and whole blood irradiation not only may use small quantities of unsealed sources and liquid solutions, but also of highly radioactive sealed radioactive sources (SRS) housed in shielded assemblies. Spent SRS are extremely hazardous as they may contain large quantities of radionuclides. Programmes to collect, consolidate, store and dispose of SRS are being developed (IAEA 2008; IAEA 2005).
Defence-related wastes tend to be simpler than those from commercial nuclear applications. Wastes derived from Pu production contain high levels of sodium, due e.g. to the need to neutralise the acidic liquor before it could be stored in the carbon steel tanks built in the early days of the US defence programme at Hanford and Savannah River. Generally, defence wastes do not contain the high concentrations of FPs found in commercial wastes, the exception being the calcined naval reactor wastes currently stored at the Idaho National Laboratory, INL but destined for the Waste Isolation Pilot Plant, WIPP, in New Mexico, USA. Donald (2007) gave generic compositions for both commercial and defence wastes (Table 2), and although there are very large compositional ranges for the constituents it does highlight the lower proportion of FPs but higher proportion of actinides present in defence waste.
Table 2. Generic compositions of typical radioactive wastes (mass%).
| Constituent | Commercial waste | Defence waste |
| Na2O | 0–39 | 0–16 |
| Fe2O3 | 2–38 | 24–35 |
| Cr2O3 | 0–2 | 0–1 |
| NiO | 0–4 | 0–3 |
| Al2O3 | 0–83 | 5–9 |
| MgO | 0–36 | 0–1 |
| MoO3 | 0–35 | 0–1 |
| ZrO2 | 0–38 | 0–13 |
| SO4 | 0–6 | 0–1 |
| NO3 | 5–25 | 0–21 |
| Fission product oxides | 3–90 | 2–10 |
| Actinide oxides | <1 | 2–23 |
| Other constituents | - | 17–27 |
In addition to the wastes generated from commercial energy supply and during the manufacture of warheads, there is also excess plutonium which has been declared surplus to requirements following the decision by the USA and Russia to reduce their warhead stockpiles. Under the 1993 Non-Proliferation and Export Control Policy the USA declared >50 tonnes of plutonium surplus to national security needs. A similar quantity was also declared surplus by Russia. These quantities may be farther increased following the 2010 USA-Russia strategic arms reduction agreement. The UK has over 100 tonnes of Pu stored at Sellafield and the Government currently plans to utilise this where possible in MOX fuel although some contaminated Pu may be immobilised in ceramic wasteforms and the possibility of burning in CANDU/PRISM reactors is being considered.
Radioactive contamination and waste may also arise from accidents. Accidents generate radioactive waste of volume and composition which depend on the material involved and the magnitude of the accident. The International Nuclear and Radiological Event Scale (INES) was developed in 1990 by international experts convened by the IAEA and the OECD Nuclear Energy Agency (OECD/NEA) with the aim of communicating the safety significance of events at nuclear installations (IAEA 2008a). The INES facilitates understanding, using a numerical rating to explain the significance of nuclear or radiological events in a similar fashion to the Richter scale for earthquakes. INES applies to any event associated with the transport, storage and use of radioactive material and radiation sources. Such events can include industrial and medical uses of radiation sources, operations at nuclear facilities, or the transport of radioactive material. Events are classified at seven levels: Levels 1–3 are “incidents” and Levels 4–7 “accidents”. These levels consider three areas of impact: people and the environment, radiological barriers and control, and defence in depth. The scale is designed so that the severity of an event is about ten times greater for each increase in level on the scale. Events without safety significance are called “deviations” and are classified Below Scale/Level 0. The partial core meltdown accident at Three Mile Island (TMI), Pennsylvania, USA in 1979 was at level 5 on the INES scale as was the UKs Windscale pile fire of 1957 while those at Chernobyl and Fukushima were level 7.
The Windscale accident occurred when the core of the Unit 1 nuclear reactor, now Sellafield, caught fire, releasing substantial amounts of radioactive contamination into the surrounding area. The fire burned for three days and radioactive material released spread across Europe. Of particular concern was release of 131I which may lead to cancer of the thyroid, and it has been estimated that the incident caused 240 additional cancer cases. No one was evacuated from the surrounding area, but there was concern that milk might be contaminated. Milk from about 500 km2 of nearby countryside was diluted and destroyed for about a month. A 2010 study of workers directly involved in the cleanup found no significant long term health effects from their involvement. Proper management of the TMI accident meant that there were no person overexposures to radiation and no casualties, so keeping it at level 5.
The Chernobyl accident in 1986 was due to lack of care in operation and disregard for standard safety procedures. The resulting steam explosion and fire released about 5% of the radioactive reactor core into the atmosphere. Some 31 people were killed in the first few weeks after the accident, 3 from blast injuries and a further 28 from high doses (>4 Sv) of radiation. Although some 6000 cancers in children and young adults have been attributed to exposure to radioiodine in fallout, only 15 patients have died from thyroid cancer in 25 years. An authoritative UN report in 2008 (UN 2011) concluded that there is no scientific evidence of significant radiation-related health effects, other than thyroid cancer, to most people exposed to radiation during or after the accident. The major effect was psychological due to the fear of radiation, rather than the health effects of the radiation itself.
The most recent accident was that of March 11th 2011 at Fukushima in Japan. A major earthquake, followed by a 15 m tsunami caused the deaths of over 20,000 people and led directly to the shutdown of three reactors and eventually to significant escape of radioactive material to the environment. Three of the Fukushima Daiichi reactors cores were severely damaged in the first three days, releasing high levels of radioactive materials into the land, sea and air environment. The Japanese authorities announced an official ‘cold shutdown condition’ in mid-December, as reactor temperatures had fallen to below 80°C at the end of October 2011. According to the Japanese government, the total amount of radioactivity released to date is approximately one-tenth that released during the Chernobyl disaster. However, the full extent and level of radioactive contamination remain unclear. Reports of the radiation doses received by the population, particularly children, suggest that these may be much lower than expected, as a result of the prompt measures taken by the authorities to evacuate the population and cut the consumption of contaminated food (Matsuda et al. 2013). It is therefore unlikely that there will be a significant rise in cancer due to exposure from this accident (Wakeford 2011). However, the psychological effects on the health of the population following the accident should not be underestimated (Brumfiel 2013).
Mechanisms for managing controlled radioactive wastes are invariably under national government control with legislative and regulatory systems in place to ensure safety and security. In the UK, for example, in 2004 the Government commissioned an independent Committee on Radioactive Waste Management (CoRWM) and in 2005 it established the NDA to ensure its 20 civil public sector nuclear sites were decommissioned and cleaned up, safely, securely, cost effectively in ways that would protect the environment for this and future generations. CoRWM recommended to Government (CoRWM 2006) that geological disposal be the end point for long-term management of radioactive wastes but with robust storage in the interim period with provision against delay or failure in reaching the end point. It also recommended a staged process with flexibility in decision making and partnership with communities willing to participate in the siting process and an expanded national R&D programme to support the process. In response the Government published a White Paper outlining the process and stages that would lead to permanent geological disposal of the UKs wastes (DEFRA 2008).
An invitation was sent out to communities in stage 1 inviting expressions of interest in hosting a repository or geological disposal facility (GDF). In stage 2 simple criteria were used to determine if the location was likely to be suitable. At this stage areas were ruled out e.g. if they had mineral resources which might prove useful in future or aquifers. Communities in potentially suitable areas could decide to participate further in stage 3 while in stage 4 desk-based studies would be carried out which would lead to borehole investigations in stage 5 prior to actual construction of the GDF underground in stage 6. Extensive work is needed during the early stages to underpin the safety case to the regulators to allow construction and safe operation and eventual closure of the GDF including decades of R&D. This volunteer approach also needs intensive public and stakeholder engagement to convince communities that this is the right approach to dealing with the waste problem. UK Government extended the NDAs responsibility to include geological disposal of the waste and in 2007 it established the Radioactive Waste Management Directorate (RWMD) as the implementing body responsible for constructing the GDF. Three communities in the Sellafield area expressed interest in participating in the MRWS volunteer process in 2007/08 but after extensive efforts withdrew early in 2013. The UK Government remains committed to volunteerism and is examining options for a new, perhaps simplified, process.
While the above highlights the need for a clear end point (permanent geological disposal), political will and public support, much radioactive waste management must be done prior to disposal. Radioactive waste management approaches vary from country to country. However, a key aspect is to know what waste you have. A national inventory must be collected as is done in the UK (NDA 2010) and all other countries. Disposal is the final step in managing radioactive wastes whereas predisposal includes activities such as decommissioning, pre-treatment, treatment, conditioning, immobilisation, storage and transport (Ojovan and Lee 2005). While various disposal options are available it is most likely that immobilised wastes will be disposed of in GDF’s of one sort or another. Waste management requires a series of steps:
pursuing opportunities for waste minimisation re-use and recycling waste treatment packaging storage transport and then final disposal where required.Waste minimisation is a process of reducing the amount and activity of waste materials to a level as low as reasonably achievable. Waste minimisation is now applied at all stages of nuclear processing from power plant design through operation to decommissioning. It consists of reducing waste generation as well as recycling, reuse and treatment, with due consideration for both primary wastes from the original nuclear cycle and secondary wastes generated by reprocessing and clean-up operations. Waste minimisation programmes were largely deployed in the 1970s and 80s. The largest volume of radioactive waste from nuclear power production is LLW. Waste minimization programmes have achieved a remarkable tenfold decrease of LLW generation over the past 20 years, reducing LLW volumes to ~ 100 m3 annually per 1 GW(e).
Recycling means recovery and reprocessing of waste materials for use in new products. Recycled waste can be substituted for raw materials reducing the quantities of wastes for disposal as well as potential pollution of air, water, and land resulting from mineral extraction and waste disposal. However, recycling has certain limitations when applied to radioactive materials. Due to their inherent radiation radionuclides are much more difficult to recover from contaminated materials. Recovery usually presumes concentration of species into a smaller volume even though this may result in more dangerous materials. Waste radionuclides recovered from contaminated materials are difficult to recycle in new devices or compounds. Hence even materials which contain large amounts of radioactive constituents (e.g. SRS) often are immobilised (conditioned) and safely stored and disposed of rather than recycled.
One example of recycling in the nuclear industry is of spent fuel. There are 435 currently operating NPP’s in 30 countries which produce 368.2 GWe. A typical NPP generating 1 GW(e) produces annually approximately 30 t of SF. The annual discharges of spent fuel from the world’s power reactors total about 10,500 tonnes of heavy metal (t HM) per year and the total amount of spent fuel that has been discharged globally is approximately 334,500 tHM (Bychkov 2012). During use only a fraction of fuel is burnt generating electricity but also forming transmutation products that may poison it. After use, the fuel elements may be placed in storage facilities with a view to permanent disposal or be reprocessed to recycle their reusable U/and Pu. Most of the radionuclides generated by the production of nuclear power remain confined within the sealed fuel elements. Currently only a fraction of SF is reprocessed in countries such as France and the UK although countries with large nuclear power programmes such as Russia and China plan to significantly increase the reprocessing capacity (Table 3). Also the US is reviewing the approach to open nuclear fuel cycle considering reprocessing as a viable option.
Table 3. SF recycling capacities, tonnes per year (Bychkov 2012).
Recycling of fissile elements (U, Pu) from SF, despite the complexity of such a process, results in a significant reduction of toxicity of the radioactive wastes (Fig. 1).
Figure 1. Relative radiotoxicity of SF and resulting HLW on reprocessing and recycling. FP-fission products, MA – minor actinides. The time required to achieve the initial toxicity of uranium ore is significantly reduced on recycling and transmutation of MA.
Another potential example of recycling in the nuclear industry is of military grade Pu much of which is stockpiled in the USA, Russia and the UK; a legacy of the cold war. Since 1972 world production of plutonium has exceeded demand for all purposes. The total world plutonium inventory is not reported but a rough calculation indicates at least 2000 metric tonnes at the beginning of the 21st century. It is technically possible to convert this material into a mixed U/Pu oxide (MOX) reactor fuel so that it can be used to generate energy in a suitable nuclear reactor. MOX nuclear fuel consists either of UO2 and PuO2 either as two phases or as a single phase solid solution (U,Pu)O2 (Burakov et al., 2010). The content of PuO2 may vary from 1.5 wt.% to 25–30 wt.% depending on the type of nuclear reactor. Whereas most efficient burning of plutonium in MOX can only be achieved in fast reactors it is currently used in thermal reactors to provide energy although the content of unburnt plutonium in spent MOX fuel remains significant (> 50 %).
Key aspects of waste management are to reduce the hazards associated with wastes and the volume of the waste material. Hazard can be reduced substantially by converting highly mobile liquid or gaseous wastes into stable solid forms.
Choosing a suitable wasteform for nuclear waste immobilisation is difficult and durability is not the sole criterion. In any immobilisation process where radioactive materials are used, the process and operational conditions can become complicated, particularly if operated remotely and equipment maintenance is required. Therefore priority is given to reliable, simple, rugged technologies and equipment, which may have advantages over complex or sensitive equipment. A variety of matrix materials and techniques is available for immobilisation (NRC 2011). The choice of the immobilisation technology depends on the physical and chemical nature of the waste and the acceptance criteria for the long-term storage and disposal facility to which the waste will be consigned. A host of regulatory, process and product requirements has led to the investigation and adoption of a variety of matrices and technologies for waste immobilisation. The main immobilisation technologies that are available commercially and have been demonstrated to be viable are cementation, bituminisation and vitrification. Immobilisation can be simply physically surrounding the waste in a barrier material (largely the case in cementation) or chemically incorporating it into the structure of a host material (largely the case in vitrification).
Cementation uses hydraulic cements to physically surround solid ILW that is contained in steel drums (Glasser 2011). Ordinary Portland cement (OPC) is the most common type of cement used for immobilizing liquid and wet solid wastes worldwide. Several OPC based mixtures are currently used to improve the characteristics of wasteforms and overcome the incompatibility problems associated with the chemical composition of certain types of radioactive waste. Composite cement systems may use additional powders as well as OPC such as Blast Furnace Slag (BFS) and Pulverised Fuel Ash (PFA). These offer cost reduction, energy saving and potentially superior long-term performance. As well as the wasteform matrix OPC’s will be used in structural components of any GDF (such as walls and floors) and are potential backfill materials so an understanding of their durability in an underground environment even without waste is important.
Embedding radioactive waste in bitumen has been used in immobilisation since the 1960’s and the total volume of radioactive waste immobilised in bitumen currently exceeds 200,000 m3. In the bituminisation process, radioactive wastes are embedded in molten bitumen and encapsulated when the bitumen cools. Bituminisation combines heated bitumen and a concentrate of the waste material in either a heated thin film evaporator or extruder containing screws that mix the bitumen and waste. The waste is usually in the form of a slurry, for example salt aqueous concentrates or wet ion exchange resins. Water is evaporated from the mixture to about 0.5% moisture, intermixed with bitumen so that the final product is a homogeneous mixture of solids and bitumen, termed bitumen compound. Its retention properties usually exceed those of cements at higher waste loadings. Bituminisation is particularly suitable for water-soluble radioactive wastes such as bottom residues from evaporation treatment and spent organic ion exchangers. However, a drawback of bitumen is its potential fire hazard. The possibility of combustion in the case of an accidental fire has led to certain restrictions on the use of bitumen as an immobilising matrix.
Vitrification is an attractive immobilisation technique because of the small volume of the resulting wasteform, the large number of elements which can be incorporated in it and its high durability. The high chemical resistance of glass allows it to remain stable in corrosive environments for long periods. Waste vitrification technology is a compromise between the desired durability of the final wasteform and its processing efficiency (Ojovan and Lee 2005, Jantzen 2011, Donald 2010). The most durable materials would require very high processing temperatures (>1500°C) which cannot be used because at high temperatures waste radionuclides occur in volatile species, generating large amounts of secondary wastes and diminishing the immobilisation efficiency. The most common glasses used in vitrification of nuclear waste are borosilicates and phosphates. Vitrification has been used for nuclear waste immobilisation for more than 40 years in France, Germany, Belgium, Russia, UK, Japan and the USA. The total production of all vitrification plants is >28,250 tonnes (Jantzen 2011, Gin et.al. 2013). Vitrification is also currently used for immobilisation of LILW.
The highest degree of volume reduction and safety is achieved through vitrification although this is the most complex and expensive method requiring a relatively high initial capital investment. However, difficult legacy waste streams are known for which current technology is inadequate, so that new approaches must be developed. These comprise development of new waste forms such as crystalline ceramic and composite radionuclide hosts as well as of new immobilising technologies such as thermochemical and in situ methods (Lee et al., 2013b). New approaches aim also to create geochemically-stable materials in equilibrium with the disposal environment to ensure a safer nuclear waste disposal scenario.
Glass composite materials (GCM’s) are used to immobilise glass-immiscible waste components such as sulphates, chlorides, molybdates and refractory materials requiring unacceptably high melting temperatures. GCM’s comprise both vitreous and crystalline components (Lee et al. 2006). Depending on the intended application, the major component may be a crystalline phase with a vitreous phase acting as a bonding agent, or, alternatively, the vitreous phase may be the major component, with particles of a crystalline phase dispersed in the glass matrix. GCM’s may be produced by dispersing both melted materials and fine crystalline particles in a glass melt and may be used to immobilise long-lived radionuclides (such as actinide species) by incorporating them into the more durable crystalline phases, whereas the short-lived radionuclides may be accommodated in the less durable vitreous phase. GCMs may also be glass ceramics where a glass is crystallised in a separate heat treatment step (Caurant et al. 2009), The French have developed a U-Mo GCM to immobilise Mo-rich HLW. Another example is the GCM developed to immobilise sulphur-enriched waste streams in Russia containing conventional borosilicate glass vitreous phase with uniformly distributed particles comprising up to 15% by volume of yellow phase (Sobolev et al. 2005).
GCMs are being developed in many countries for their difficult wastes. E.g. the UK’s most hazardous wastes are those in the legacy ponds and silos (LP&S) at Sellafield. A number of novel thermal technologies are being examined to immobilise the complex, often ill-defined and heterogeneous wastes found in the LP&S. These include pyrolysis steam reforming, plasma vitrification and Joule heating in container melting (JHCM). In the latter process mixed solids and sludge wastes are placed in a concrete lined steel container with embedded graphite electrodes in the comer and melted to produce a stable solid.
While JHCM can successfully convert reactive material (e.g. metals, sludges & organics) to more stable forms the variable nature of the wastes make control of process and product difficult and it is difficult to characterise both the heterogeneous waste and product. Much R&D is needed including durability testing of the products of these technologies. However, their use has seriously reduced the hazard from the original wastes and pragmatic engineering approaches such as these are needed even if the resulting wasteform is not as perfect as ultimately desirable.
Single phase ceramics such as zircon (ZrSiO4) can potentially host a large number of nuclides and can be used as a monophasic wasteform. However, monophase ceramics are difficult to fabricate and polyphase compositions are more common. The composition of the polyphase ceramic can host multiple radionuclides and be tailored to that of the waste composition to achieve complete and reliable immobilisation of the waste constituents. The most famous polyphase ceramic for nuclear waste immobilisation is Synroc. Synroc is short for “Synthetic Rock”, invented in 1978 by T. Ringwood of the Australian National University. Synroc is made of geochemically-stable natural titanate minerals which have immobilised uranium and thorium for billions of years. U/Th-containing natural analogues of the basic constituent of Synroc - zirconolites from Sri Lanka dating back 550 million years while amorphized have nonetheless withstood the alteration processes of their natural environment.
When examining storage options for radioactive waste it is important to consider the whole storage system rather than concentrating on just the store building itself (CoRWM 2009). A number of interacting components and operations combine and contribute to create the necessary robust, safe and secure storage arrangements. These factors must be considered in an integrated manner. There are two main concepts in storage of radioactive waste. If the packaged wasteforms are basic then a high quality often shielded store will be needed. On the other hand if the wasteform is high quality and shielded then the store can be of poorer quality or the waste containers can simply be left in the open. A generic shielded store is shown in Figure 2 and an example of a high quality store has recently been constructed at Hunterston in Scotland (Figure 4) which has 2 metre thick reinforced concrete walls and roof and careful control of atmosphere.
Figure 2. Schematic of a UK generic shielded store.
The wasteform or product, its container, the building structure, the ventilation system, the handling equipment, the monitoring and inspection regime and the maintenance and refurbishment regime all have roles to play in ensuring safety and security of the store. The waste storage system involves a number of levels (NDA 2010b). The wasteform (1) is the primary protective barrier, the waste container (2) is the secondary barrier. Control of store environment (3) is important in maintaining the integrity of the wasteform and waste container while the store structure (4) is the final layer of weather/atmosphere protection for waste package and an important element of physical security of waste.
Packages inevitably evolve during storage and those changes affecting the safety function need to be understood and controlled to satisfy the regulators of the safety of the store and waste. Different storage concepts and designs require different performances from these various components and operations and therefore place different degrees of reliance on them. Quite different combinations of them can provide equally safe and secure storage. For example, most existing modern stores in the UK have massive concrete structures holding unshielded containers, but the alternative “mini-store” concepts rely on heavily shielded containers within lightly built stores. This latter concept is used in some other EU countries. In a storage system not every component need last for the whole design life. It is possible at the design stage to plan to replace or refurbish various components and build in at the outset specific features to enable this. More straightforward items to consider are building fabrics, external ventilation systems and power supplies. The more complex refurbishments or replacements to consider are cranes, active area surveillance equipment and major building structures.
In the UK options being examined for SF include Multi Purpose Containers (MPC) suitable for storage, transport & disposal of a range of SF types (Figure 3A).
Figure 3. A: MPC options under development for UK SF. B: Yellow boxes for storing ILW.
As well as high quality packages for SF they have also been developed for less active wastes. So-called yellow boxes (Figure 3B) have been used extensively in Europe and used to store spent resin waste from existing storage tanks at the Dungeness plant in England. The containers are transportable and offer self-shielded protection, weighing around 18 tonnes when empty. The waste is expected to be stored in them for at least a decade.
Waste is continuously being generated and stored but in many countries without an end point of geological disposal in sight an issue is whether to store all waste at the sites where it is generated or to consolidate it at centralised national or regional stores. The Blue Ribbon Commission (BRC) e.g. recommended this option be examined in the USA and it is also being considered in the UK.
Options for disposal depend to large extent on the content and half-life of radionuclides in the waste. Small contents and short-lived wastes may be suitable for near surface disposal (IAEA 2002) while larger contents and long-lived radionuclides require deep or very deep disposal relying on the geosphere to keep the radioactive species from the biosphere (IAEA 2003, Ahn, Apted 2010). As for the storage concepts described above most geological approaches use a multi-barrier system to improve the safety of disposal where the wasteform, container, near field environment (e.g. engineered barrier system, EBS) and far field environment (host rock) are all important in retaining radionuclides in the geosphere.
Near surface disposal sites are constructed anywhere from on the surface itself to up to 60m below it. Such facilities with an EBS are suitable for most LLW and LILW and are widespread across Europe (e.g. Drigg in the UK, El Cabril in Spain) and also are used in the USA and Japan. Globally we have decades of experience of operating such disposal sites. The EBS, which typically consist of clay or other barrier layers, is necessary to reduce radionuclide leach rates from the waste and to divert water away from the wastes. Water management is used during the operational phase of these facilities when the waste packages are uncovered as water cannot be allowed to accumulate within the waste cells.
Geological disposal in a mined repository is the most likely option for HLW, SF, SRS and long-lived LILW (IAEA 2003, Ahn, Apted 2010). The main concepts of geological disposal are wet and dry, typically at depths from 500–1500 m. The wet option is a mined and engineered repository located so that eventual water ingress and saturation is inevitable. Various types of host rock are being considered governed largely by the local geology including hard rock (e.g. granite as in the Swedish and Chinese concepts) and soft rocks (e.g. clays in France and Belgium). The dry mined and engineered repository concept was favoured in the USA including high and dry (Yucca Mountain, Nevada) and shallow and dry (the Waste Isolation Pilot Plant [WIPP] located in salt in Carlsbad, New Mexico. However, technical (and other) problems at Yucca Mountain including that it was not as dry as hoped have led to the demise of that programme.
The UK concept is for multiple vaults in the same region (Figure 5A) to accommodate the complex array of wasteforms that we have, a legacy of our early indecision on which reactor type to build and of military and research programmes. A clear research need is to understand wasteform evolution during storage and disposal and the eventual interaction of the corrosion products from the different parts of the GDF.
Figure 5. A: The UKs current multi-vault GDF concept. B: Schematic of very deep borehole disposal.
Another concept is that of very deep (permanent) disposal. In this concept (Gibb 2000) the waste is located at depths of 3 km or more and as such any transport of radionuclides through the geosphere is extremely limited (Figure 5B). Further, if located in suitable (granitic) rock the radiogenic heat from HLW can cause reaction with the surrounding rock and lead to creation of a sarcophagus or granite coffin which seals the waste in permanently. The US BRC (2012) was positive about the deep borehole disposal concept and the US is planning a demonstration programme. However, this is untried technology needing safety case that potentially could take many decades to come to fruition.
The UKs waste management programme has been described in a global context highlighting options for immobilisation, storage and disposal.
Ann J., Apted M.J. 2010. Geological repository systems for safe disposal of spent nuclear fuels and radioactive waste. Woodhead, Cambridge, 792 p..
BRC 2012, Report to the Secretary of Energy, Blue Ribbon Commission on American’s Nuclear Future, January.
Brumfiel G. 2013. Fallout of Fear. Nature 493 (7432) 290–293. doi: 10.1038/493290a.
Burakov B.E., Ojovan M.I. and Lee W.E. 2010. Crystalline Materials for Actinide Immobilisation, Imperial College Press, London, 198 pp..
Bychkov A.V. 2012. Status and prospects of advance nuclear fuel cycles. INPRO Dialogue Forum. July 30–August 3, 2012, IAEA, Vienna.
Caurant D., Loiseau P., Majérus O., Aubin Chevaldonnet V., Bardez I. and Quintas A. 2009, Classes, Glass-Ceramics and Ceramics for Immobilization of Highly Radioactive Nuclear Wastes. Nova, New York.
CoRWM 2006. Managing our Radioactive Waste Safely (Department of the Environment and Rural Affairs, DEFRA, July).
CoRWM 2009. Interim Storage of Higher Activity Wastes and the Management of Spent Fuels, Plutonium and Uranium. (CoRWM doc. 2500, DECC).
DEFRA 2008. Managing Radioactive Waste Safely. A Framework for Implementing Geological Disposal (DEFRA June).
Donald I.W., 2007, Immobilisation of Radioactive and Non-radioactive Wastes in Glass-based Systems: an Overview, Glass Technol: Eur. J. Glass Sci. Technol. A, 48, 155–163.
Donald I.W. 2010, Waste Immobilisation in Glass and Ceramic Based Hosts. Wiley, Chichester.
Gibb, F.G.F. 2000 A new scheme for the very deep geological disposal of high level radioactive waste, J. Geol. Soc. (London) 157, 27–36.
Gin S., Abdelouas A., Criscenti L.J. et.al., An international initiative on long-term behavior of high-level nuclear waste glass. Materials Today, 16 (6), 243–248.
Glasser F. 2011. Application of inorganic cements to the conditioning and immobilisation of radioactive wastes. In: M I Ojovan.. Handbook of advanced radioactive waste conditioning technologies, 67–135, Woodhead, Cambridge.
IAEA 2002. Scientific and Technical Basis for the Near Surface Disposal of Low and Intermediate Level Waste, Technical Reports Series No. 412, IAEA, Vienna.
IAEA 2003. Scientific and Technical Basis for the Geological Disposal of Radioactive Wastes, Technical Reports Series No. 413, IAEA, Vienna.
IAEA 2005. Disposal Options for Disused Radioactive Sources, Technical Reports Series No. 436, IAEA, Vienna.
IAEA 2008. Locating and Characterizing Disused Sealed Radioactive Sources in Historical Waste, IAEA Nuclear Energy Series No. NW-T-1.17, IAEA, Vienna.
IAEA 2008a. INES, the International Nuclear and Radiological Event Scale User’s Manual. 2008 Edition. IAEA, Vienna.
IAEA 2009. Classification of radioactive waste: General Safety Guide GSG-1. IAEA, Vienna.
Jantzen C.M. 2011. Development of glass matrices for high level radioactive wastes, In: Handbook of advanced radioactive waste conditioning technologies, Ed. M.I. Ojovan. pp. 230–292. Woodhead, Oxford.
Lee W.E., Ojovan M.I., Stennett M.C., Hyatt N.C. 2006. Immobilisation of Radioactive Wastes in Glasses, Glass Composite Materials and Ceramics, Adv. Applied Ceramics 105 [1] 3–12.
Lee W.E., Ojovan M.I. and Jantzen C.M. 2013a (Editors), Radioactive Waste Management and Contaminated Site Clean-up: Processes, Technologies and International Experience, Woodhead, Cambridge UK.
Lee W.E, Gilbert M., Murphy S. and Grimes R.W. 2013b, Opportunities for Advanced Ceramics and Composites in the Nuclear Sector, J. Am. Ceram. Soc. 96 [7] 2005–30 (2013).
Matsuda N., Kumagai A., Ohtsuru A., Morita N., Miura M., Yoshida M., Kudo T., Takamura N., Yamashita S. (2013). Assessment of internal exposure doses in Fukushima by a whole body counter within one month after the nuclear power plant accident. Radial Res. 179 (6): 663–668 doi: 10.1667/RR3232.1.
NDA 2010. The 2010 UK Radioactive Waste Inventory.
NDA 2010a. Geological disposal. Generic environmental safety case. Main report. Report NDA/RWMD/021.
NDA 2010b. Package Evolution Status Report of the generic Disposal System Safety Case.
NDA 2012. Exotic Fuel and Nuclear Materials – Dounreay, Credible Options, NDA UK Feb.
NRC 2011. Waste Forms Technology and Performance: Final Report. National Research Council. 340 p., The National Academies Press, Washington, D.C..
Ojovan M.I., Lee W.E. 2005. An Introduction to Nuclear Waste Immobilisation. Elsevier, Amsterdam.
Sobolev I.A., Dmitriev S.A., Lifanov F.A., Kobelev A.P., Stefanovsky S.V., Ojovan M.I. 2005. Vitrification processes for low, intermediate radioactive and mixed wastes. Glass Technology, 46, 28–35.
UN 2011. Sources and Effects of Ionising Radiation. UNSCEAR 2008 Report to the General Assembly with Scientific Annexes. Vol. II. United nations, New York. http://www.unscear.ora/docs/reports/2008/11-80076_Report_2008_Annex_D.pdf (2013).
Wakeford R.J. 2011. And now, Fukushima. Radiol Prot. 31 (2) 167–176. doi: 10.1088/0952-4746/31/2/E02.