This was the last research fast reactor in the research programme. The next reactor would have been a commercial size demonstration fast reactor. Con­struction of PFR started in 1968 and it operated from 1974 to 1994 with an output of 250 MW. Although the reactor was not of commercial size, the PFR fuel assemblies were designed and manufactured at commercial size. It was cooled by sodium as enough experience and confidence had been gained from DFR operations to go forward without the need for a eutectic coolant.

After shutdown, the fuel was removed to the associated fuel pond where it remains until a decision on its final treatment and destination is made.

The experience gained from the early DFR NaK decommissioning enabled the world’ s largest sodium destruction plant to be designed and built during the late 1990s. The plant was built in the decommissioned and stripped out turbine hall. Around 1,500 tonnes of primary and secondary sodium were destroyed by the same process as that used for DFR. The dif­ference from DFR was that the sodium was only lightly contaminated as PFR fuel was in sealed elements whereas DFR fuel was in vented elements so reaction products contaminated the NaK.

As with DFR, there is a current programme for cleaning up the residual sodium and decommissioning the pipework, vessels and reactor compo­nents over the next decade. The additional challenge of decommissioning the PFR reactor itself is that it is 15 m below ground level and this requires the design and operation of special remote handling appliances.

Decisions in the UK and Scotland within the nuclear industry and with regulators on the concepts and practicability of in-situ disposal will have a bearing on the eventual end state of the decommissioned PFR reactor.

18.1.4 Spent fuel and RAW classification

The US classification system has two separate subsystems: one applies to commercial waste, and NRC regulations define it; the other applies to DOE

SNF and waste. The two systems are used for different purposes and differ­ent situations so conflicts do not occur. If ownership of radioactive waste is transferred from the DOE to a commercial entity licensed by the NRC, the waste is then subject to NRC regulation (and classification).

Spent fuel

The United States defines ‘SNF’ as fuel that has been withdrawn from a nuclear reactor following irradiation, the constituent elements of which have not been separated by reprocessing. US law generally uses the term ‘SNF’ rather than ‘spent fuel,’ and the DOE has begun using the term ‘used fuel’ to acknowledge that in the future, the material may have residual value through recycling. For the purposes of this chapter, used fuel is referred to as SNF in accordance with the conventional terminology unless otherwise noted.

Used fuel

There are 22 CANDU[31] (CANada Deuterium-Uranium) power reactors in Canada owned by three provincial electric utilities. Ontario Power Genera­tion Inc. (OPG) owns 20 reactors (eight of which are leased to Bruce Power Inc. for commercial electricity production), while Hydro-Quebec (HQ) and New Brunswick Power (NBP) each own one reactor. All CANDU® fuel bundles are fabricated from natural uranium oxide pellets, contained in a zirconium-alloy (Zircaloy-4) sheath. The weight of a nominal bundle is 23.6 kg, of which 21.3 kg is due to the uranium oxide, approximately 19.2 kg can be attributed to the uranium (without the oxygen component). Each year, 4,500-5,400 used fuel bundles are generated per reactor, based on 80-95% full power reactor operation (CNSC, 2008). A 600MW CANDU® nuclear reactor produces approximately 20 m3 of used nuclear fuel per year.

Nuclear fuel wastes from nuclear power generating stations are stored in wet and dry states at the locations where they are generated. The used fuel

is first placed in water-filled fuel storage bays, and after several years (i. e., six to ten years) the used fuel can be transferred to an on-site dry storage facility. These dry storage facilities are large, reinforced concrete cylinders or containers. Each nuclear power generating station in Canada has enough storage space to store all the used fuel produced during the operating life of the station.

Following a decade-long environmental assessment of a deep geological disposal concept for nuclear fuel waste that ended in 1998, the Government of Canada passed the NFWA in 2002, which made owners of nuclear fuel waste responsible for the development of long-term waste management approaches. Shortly after the NFWA came into force, the nuclear energy corporations, OPG, HQ and NBP, established the NWMO and each waste owner established trust funds to finance the implementation of long-term waste management activities. The NWMO’s mandate is to explore options for the long-term management of Canada ’s nuclear fuel waste, provide proposals to the Government of Canada and to implement the selected approach.

Following extensive studies and public consultation, the NWMO pre­sented four options, including those listed in the NFWA, namely long-term storage at the reactor sites, central shallow or below ground storage, deep geological disposal, and lastly an option called the adaptive phased manage­ment (APM) approach (NWMO, 2005). The APM approach essentially combines the three above listed options within a flexible adaptive manage­ment decision-making process. In 2007, the Government of Canada announced that it had selected the APM approach for the long-term man­agement of used fuel in Canada. With this government decision, NWMO assumed responsibility for implementing the APM approach (NWMO, www. nwmo. ca).

1. Only LLW and ILW radioactive waste that has been conditioned into a solid shall be accepted at Vaalputs disposal repository. This includes solidified waste streams such as sediments, resins and sludge.

2. Radioactive waste shall be safely managed in a regulated manner, compatible with internationally and nationally agreed principles and standards.

3. The waste generator shall demonstrate that the waste packages comply with the Necsa quality standards for the manufacture of containers and the filling and storage of those waste packages [11].

4. PDO will have the right to perform audits on the waste generators’ waste management system and the implementation thereof.

The uranium conversion facility (UCF) located at KAERI was operated from 1982 to 1992. After the localization of nuclear fuel fabrication technol­ogy, it was shut down in 1993. UCF decommissioning began in 2001 and radioactive waste from UCF has been stored in a temporary storage build­ing in the conversion facility. All the wastes are contaminated mainly with natural uranium. Currently, the dismantling of 26 out of 27 rooms at UCF has been conducted (Fig. 21.14) , including decontamination of concrete surfaces, removal of contaminated soil, and completion of treatment of sludge waste in a lagoon.

Research achievements to date are:

• development of volume reduction technology for large amounts of radi­oactive concrete wastes

• development of soil decontamination technology for remediation of nuclear sites after decommissioning

• development of melting technology for decontamination of a hundred tons of slightly contaminated metallic wastes generated from KRR-1 and 2 and UCF

• development of technologies for safe management of irradiated graph­ite arising from decommissioning of KRR-1 and 2

• development of a database system for management and data assessment from D&D activities

• development of chemical decontamination technology applicable to metal wastes contaminated with UN (uranium nitride), AUC (ammo­nium uranyl carbonate), and UO2 generated by dismantling UCF

• development of the safety assessment methodology of the decommis­sioning process

• simultaneous remote measurement of alpha/beta contamination in highly contaminated facility

• decontamination technology development

• waste treatment technology development.

Major R&D activities are now concentrated on development of the decommissioning waste reduction and recycling technology for commercial NPPs and nuclear facilities.

21.5 Conclusion

Given the scarcity of Korea ’s primary energy resources, nuclear power is vitally important as an engine of growth for the nation. Korea has followed a set of consistent policies and executed steady plans to expand nuclear power. With a significant share of nuclear power in the energy mix, the disposal of RAW and SF is looming large as a high-visibility national issue. A low-and intermediate-level waste disposal site has been selected and the facilities are currently under construction with its full operation expected in 2014. Spent fuel management has also become imminent. Although no satisfactory resolution is in sight in the foreseeable future, various options are being studied with the government ’s keen interest and full support. Korea has also designed a rigorous process for decontaminating waste materials.

23.4.1 Decommissioning strategy

The basic policy for decommissioning commercial NPPs was established by the JAEC in 1982. It states that retired commercial NPPs should be disman­tled as early as possible after shutdown and the site should be effectively re-used for next generation NPP. The Framework for Nuclear Energy Policy issued by the JAEC states that it is the operator’s own responsibility, but under government regulation, to carry out decommissioning of a nuclear facility, ensuring safety, while obtaining local communities’ understanding and cooperation.

The regulatory policy for dismantling or decommissioning reactor facili­ties has been discussed by the NSC. To ensure safety during decommission­ing of commercial NPPs, the regulation was implemented by applying existing provisions in the Reactor Regulation Law by the operators. To date, decommissioning of reactor facilities has been implemented at facili­ties such as the JPDR of the JAEA and the Tokai-1 NPP of JAPC, the development and application of dismantling technologies have progressed, and know-how for decommissioning has been accumulated. The NSC examined the idea of ideal safety regulation, based on the experience of decommissioning of nuclear facilities. It also took into consideration the features of nuclear facilities post-termination and the level of potential risks.

The Decommissioning Safety Subcommittee has investigated the appro­priate regulation systems for decommissioning, based on regulatory experi­ences of decommissioning reactor facilities. The Decommissioning Safety Subcommittee proposed the decommissioning regulations as: [38]

On the basis of such recognition, the Reactor Regulation Law was amended in 2005. A licensee applying for approval of decommissioning has to submit a decommissioning plan that describes, for example dismantling methods, radiation controls, safety assessment and the financial plans. The regulatory body approves the decommissioning plan after examining its conformity with technical standards. At the final stage of decommissioning, the licensee submits a document that describes the implementation status of dismantling, management of contaminated materials and the final distri­bution of contamination and requests the regulatory body’s confirmation. The decommissioning is completed after the regulatory body confirms that the measures for radiation hazard prevention are no longer necessary and management of contaminated materials is completed.

Perhaps the easiest material to immobilize, at least technologically, is Pu. It is possible to vitrify Pu in a variety of glass compositions and immobilize it in various ceramic-based hosts. In the latter half of the 1990s, a number of scoping studies were published (e. g., Matzke and van Geel, 1996; Gray and Kan, 1996; Wicks et al., 1996), which outlined a number of potential hosts and demonstrated practically that it was indeed possible to immobilize Pu in a variety of glass and ceramic hosts.

The largest of these studies was performed on behalf of the US DOE which identified 72 possible options (Gray, 1996a) of which five were studied in depth, three involving vitrification and two immobilization in a ceramic host. For the vitrification processes lanthanum borosilicate glass based on the original ‘Loeffler’ optical glass composition was selected as the host in preference to lead iron phosphate or alkali tin silicate compositions, which had also been considered as potential hosts (Gray, 1996b). Studies of boro — silicate glass developed for vitrifying waste arising from reprocessing nuclear fuel elements have shown that this glass can also be used, but suffers from having a low actinide solubility, i. e. <3 mass% Pu, which compares unfavourably with lanthanum borosilicate glass, which can incorporate in excess of 10 mass% PuO2 (Meaker et al., 1997; Peeler et al., 1997). Two of the vitrification proposals were similar in that the waste plutonium and any scrap material, pre-treated where necessary, would be vitrified in the lan­thanum borosilicate glass. However, while one proposal (Gray, 1996b) required a new facility to be built, the second (Gray, 1996c) made use of an existing facility at SRS. In both proposals the glass would be poured into stainless steel HLW canisters, whereas in the third proposal (Gray, 1996d) it would be cast into small metal cans, 20 of which would be carefully posi­tioned within a HLW canister. The size of HLW canisters, 0.6 m diameter x 3 m high and weighing 2,000 kg, mitigates against theft, but additional secu­rity was proposed for the first two proposals by spiking the glass with suf­ficient 137Cs to maintain a y-radiation field above 1 Gy/hr for 30-60 years. A slightly different approach was proposed for the third option in that the voidage surrounding the cans would be filled with conventional borosilicate glass containing either 137Cs or HLW. Compositions of a selection of the glasses suggested for immobilizing surplus Pu are given in Table 25.6. Only a few glasses containing radioactive constituents have been prepared, most candidate compositions were made containing non-radioactive rare earth oxides as surrogates for Pu.

Phosphate glasses have also been investigated as they tend to have a higher solubility for actinides than silicate glasses and have been used in Russia as an alternative to borosilicate glass for the immobilization of HLW. Initially they suffered from poor durability and were highly corrosive in the molten state, so they found less favour than borosilicate glasses for which non-active processing technology existed. However, their durability has improved with the development of sodium aluminium phosphate (Minaev et al., 2004; Donald et al., 2006), iron phosphate (Day et al., 1998; Mogus-Milancovic et al., 1997) and lead iron phosphate glasses (Sales and Boatner, 1988) and they now have durabilities which match or exceed the standard borosilicate glass. Further information on phosphate-based glasses for immobilizing wastes is given elsewhere (e. g., Donald, 2010; Jantzen, 2011).

The introduction of CCIM technology largely addresses the problem of refractory liner corrosion by containing the melt within a solid skull of glass produced by cooling the outside of the furnace. Corrosion of glass contact refractories used in the vitrification of RAW has been reviewed in, e. g., Bingham et al. (2011).

Although the majority of the effort went into investigating vitrification, there was significant effort put into the ceramic option employing a Synroc — based composition which was subsequently selected as the most suitable choice, with disposition to be carried out in a similar manner to that of the vitrification route, i. e., ceramic pellets encapsulated in HLW glass in steel containers. Synroc (synthetic rock) is the generic name for a group of ceramics containing varying proportions of minerals found in nature, includ­ing hollandite (BaAl2Ti2O6), perovskite (CaTiO3), zirconolite (CaZrTi2O7) and rutile (TiO2) (Ringwood et al., 1979). Synroc-D was initially developed for defence requirements and consists primarily of perovskite, zirconolite, nepheline (NaAlSiO. ) and spinel (MgAl. O4), together with a continuous intergranular glassy phase, whilst the ceramic developed for immobilizing Pu consisted mainly of zirconolite, the primary phase for incorporating actinide elements. Waste forms with Pu loadings in excess of 10 mass% have been reported to exhibit excellent durability (Jostsons et al., 1995). Mono­lithic waste forms can be produced from a mixture of waste and ceramic precursor powder using conventional ceramic processing techniques, i. e., hot pressing (HP), hot isostatic pressing (HIP) and cold-pressing followed by sintering (CPS). A HP process developed at the Australian Nuclear

Glass BaO/ SrO І-З2О3 B2O3 ai2o3 Si02 PbO p2o5 Gd203 Nd203 Pu02 Na20 Others Beference Lan-14 3.5 30.4 4.7 15.0 27.0 11.3 8.2 CeO Bamsey et al. (1994) Lan-17 3.3 16.2 3.9 22.5 22.5 8.8 20.6 8.3 CeO Bamsey et al. (1994) LaAIBSi 4.4 23.2 4.3 9.5 26.3 11.1 15.0 6.1 Sm203 0.2 Zr02 Bibler et al. (1996) LaAIBSi 2.5 12.4 11.7 21.5 29.1 8.6 12.8 1.3 Zr02 Vienna et al. (1996) LaAIBSi 2.4 8.6 11.4 20.8 28.2 7.5 8.6 11.4 Macfarlane (1998) LaAIBSi 2.5 20.3 13.0 10.0 20.0 11.7 15.4 Macfarlane (1998) FEP 1 62.0 38.0 Fe203 M og us-M і I a ncovic et al. (1997) FEP 2 85.0 15.0 Fe203 M og us-M і I a ncovic et al. (1997) NaAIPI 2.0 19.0 39.0 40.0 Donald et al. (2006) MMW 17.9 6.3 50.4 12.5 8.9 4.1 Li20 Harrison et al. (2008) LaBS 2.5 19.0 13.0 10.0 20.0 13.5 15.0 7.0 Hf02 Fox et al. (2008) Frit X

Science and Technology Organisation (ANSTO) utilizes a stainless steel collapsible bellows can into which the mixture is placed. After evacuating and sealing the can, it is cold pressed to approximately two-thirds of its original length before being hot-pressed. Using simulated wastes, ANSTO have demonstrated this process on an industrial scale by successfully pro­ducing samples up to 436 mm in diameter.

More recently, both these potential options have been dropped in favour of use of surplus Pu as a mixed oxide (MOX) fuel, in line with the Russian view of Pu as a strategic material rather than a waste, for use in either Pu breeder reactors or light water power reactors (Gong et al., 2001; IPFM, 2009). France and Germany started bilateral programmes with Russia in 1992 which demonstrated the feasibility of recycling weapons grade Pu in Russian VVER 1000 and BN 600 reactors (Seyve et al., 1999 ).

The UK government has not declared any weapons grade plutonium to be surplus, but began a public consultation in 2011 (DECC, 2011) into the long-term management of the large stock of UK-owned civilian pluto­nium, 114.8 te at December 2010 (www. hse. gov) . Studies funded by the Nuclear Decommissioning Authority (NDA) into a variety of topics includ­ing re-use as MOX fuel, the preferred option, and immobilization will permit decisions to be made on the management of the stocks. Immobiliza­tion of civilian stocks declared surplus or unsuitable for re-use in ceramic and vitreous waste forms is being investigated (Harrison et al., 2008) and will provide a significant read-over to weapons grade plutonium should a future need arise.

Other alternative disposition options have been suggested, including the use of some surplus Pu to produce 99Mo by irradiation of 239Pu for medical applications (Mushtaq, 2011), but these must generally be viewed as only suitable for dealing with very small quantities.

Further information can be found primarily in Czech, Slovak and Polish National Reports prepared under the Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management (Czech National Report, 2008, National Report of Slovak Republic, 2008, National Report of Poland, 2008). A wide range of information is included in a variety of IAEA reports. In particular, it is possible to find a lot of information on waste management systems in Eastern European countries in the reports dealing with WWER waste management system or remedia­tion of contaminated sites after uranium mining and milling activities (IAEA, 1995, 2000, 2004, 2005, 2006).

It is also possible to find summaries of information on waste manage­ment systems in Eastern European countries in European reports compiled before accession of these countries to the European Union (EC DGXI, 1999, 2000 ).

Next to the underground laboratory, a technology centre was constructed in 2009 in which are now displayed the prototypes of objects and machines that demonstrate what could be implemented in the disposal facility. In late 2009, ANDRA produced a ‘dossier’ (ANDRA, 2009) giving the status on the development of the planned repository. For the study of the location of underground facilities of the repository, it proposed an area of 30 km2 called the Area of Interest for Detailed Investigation (‘Zone d’Interet de Recher­che Approfondie’, or ZIRA), that was the result, apart from scientific con­siderations, of a dialogue with local stakeholders. In March 2010, ANDRA was authorized to conduct detailed geological investigations in this area. The project Industrial Centre of Geological Disposal (‘Centre Industriel de stockage GEOlogique’, or CIGEO) was launched.

The Harwell Nuclear Licensed Site in Oxfordshire has successfully cleaned up and delicensed part of the site. The site was a former RAF airfield before it was used for research associated with the development of nuclear power in the UK. The original licensed site was 113 hectares, containing four research reactors. A phased approach to delicensing was adopted based on the programme for decommissioning the facilities. In 1992 five hectares were delicensed and in 2006 a further seven hectares, originally containing 43 buildings, were delicensed. Ten of the buildings had been used for work involving radioactivity. The facilities were decommissioned and the land and buildings certified free of ionising radiation and available for non­nuclear development. A further five hectares, including the former site of the research reactor GLEEP (the graphite low energy experimental pile), were delicensed in 2011. In this instance, all the buildings were demolished and some concrete foundations were left. The case for delicensing a further five hectares has been submitted. Experience gained in the delicensing work was that it is important to pay attention to detail and to work with the ONR as far as possible. It is best practice to build delicensing require­ments into decommissioning and land remediation works and to keep good decommissioning and remediation records. Often it is important to demon­strate the absence of something, for example that the section of drain is not there any more.

Since 1984, a programme of monitoring for radioactive objects has been carried out on beaches in the vicinity of the Sellafield site in West Cumbria. During this programme, over 650 radioactive objects were identified and removed up to the summer of 2009, comprising particles with sizes smaller or similar to grains of sand and also contaminated pebbles and stones. These objects have a much higher activity content that can be easily distinguished from the ambient homogeneous levels of contamination on the beaches. The source of these objects is not known but there have been a number of known events in the past that have resulted in release of radioactive parti­cles into the environment, including early operation of the Windscale piles (1952 to 1957), the Windscale fire in 1957 and the beach incident in 1983. Hence the management strategy for the clean-up of the beaches has to consider a wider context than just the beaches, e. g., the terrestrial and marine environment.