Rosyth is a long established naval dockyard built between 1909 and 1915. It covers 127 ha and is located on the north side of the Firth of Forth in Fife (Fig 17.1). The dockyard became involved with nuclear operations in 1960 with the start of support services to the Royal Navy’s nuclear submarines. Some support work continued until 2003 although since 1993 the main support services have been provided at Devonport in England. In 1997 the dockyard was sold to Babcock International which now holds the nuclear site licence. The nuclear decommissioning liability is still retained by the MoD.

Enrichment and fuel fabrication facilities waste

The product from uranium recovery facilities is processed to enrich the fissile content. Tailings containing depleted uranium (DU) are a byproduct of the enrichment process. Fuel manufacturing facilities fabricate nuclear fuel assemblies for LWRs containing low-enriched uranium. This activity includes receipt, possession, storage, and transfer of special nuclear mate­rial. Other licensed activities supporting fuel manufacturing include uranium storage, scrap recovery, waste disposal, and laboratory services. Radioactive waste from these processes, which varies in type and amount, is managed within the classes described in Table 18.2.

Depending on available quantities and long-term and short-term needs, DU could be a resource for a variety of applications and uses, in which case it is considered source material. If DU is not a resource, the NRC catego­rizes it as Class A LLW. When 10 CFR Part 61 was developed, the disposal of large quantities of DU was not anticipated. However, with the recent licensing of fuel enrichment facilities, which will produce large quantities of DU waste, NRC determined it appropriate to revisit the issue. Therefore, NRC is examining whether the disposal of large quantities of DU from enrichment plants warrants additional, site-specific disposal protections to ensure long-term safety. As an interim measure, the NRC has issued interim guidance to states that regulate the disposal of large quantities of DU (NRC, 2010).

The DOE and private corporations (e. g., United States Enrichment Cor­poration) currently possess and store DU. The DOE manages a large stock of DU at two gaseous diffusion enrichment plants, and continues to manage it as source material available for reuse. If a decision is made that this mate­rial has no potential use, it can be disposed of in DOE or commercial LLW disposal facilities, provided the waste meets the disposal facility ’s waste acceptance requirements. Some DOE DU has been disposed of as LLW at the Nevada National Security Site (NNSS), formerly the Nevada Test Site.

AECL’s CRL site is large (-4000 ha) and diverse and contains many struc­tures and features, some dating back to the beginning of the site ’s first establishment in 1944. The site is expected to continue in operation as a licensed nuclear site, with a wide range of nuclear research and develop­ment and operation activities being conducted for many years to come. Any contaminated facilities or contaminated lands and radioactive wastes that have been produced during prior operations or decommissioning activities constitute the legacy liabilities that are now managed through the NLLP, described above in Section 19.2.2.

The decommissioning model for the CRL site, including the waste man­agement areas, is described in a comprehensive preliminary decommission­ing plan (CPDP) (Miller, 2010). The strategy developed is for a number of individual decommissioning projects for the site’s various components over time rather than a single project for the site as a whole at some time in the future, designated as the end of operational life. Priorities for decommis­sioning projects are established based on health, safety, security and envi­ronmental risks, and also take into consideration operational requirements and business priorities (Stephens, 2009). The individual decommissioning projects in the CRL site CPDP document are grouped into seven planning envelopes (PE), where each PE is a grouping that has a degree of similarity, which lends itself to the application of common planning assumptions. Plan­ning envelopes 1-4 are for above-ground structures, PE 5 is for distributed services, PE 6 is for affected lands, and PE 7 is for waste management areas. The individual projects will, in general, take each respective structure or feature to a documented end-state while the site as a whole continues in operation. Some projects will be implemented at the end of the site’s opera­tional life to qualify the site for a period of institutional control, the refer­ence being 300 years (2100-2400). During the institutional control period, selected parts of the site may be turned over for industrial re-use in accord­ance with then-current laws and regulations. Work is currently under way to develop an overall, co-ordinated environmental restoration strategy for the CRL site which is integrated with the plans for decommissioning the physical structures and available waste management facilities, and which is in the most cost — and risk — effective manner as determined by the various stakeholders.

The national radioactive waste management policy and strategy document [3] prescribes the use of a balanced and systematic way of evaluating respec­tive waste management options using a multi-attribute analysis approach such as the BATNEEC process. It was, however, decided to utilize the more recent best practical environmental option (BPEO) and best practicable means (BPM) multi-attribute analysis processes. The BPEO process will be utilized for the selection of the management options whilst the BPM process will be followed for the refinement of design and operational conditions.

BPEO studies are particularly relevant to strategic decision making, involving choices between alternative management options. The fundamen­tal comparison relates to the performance of environmental options, but the process should provide a holistic appraisal, which includes continual improvement, of factors associated with the practicability of implementing strategic alternatives. BPM relates to optimization of the selected option from the perspective of radiological protection, and is concerned with the detailed refinement of design and operation conditions.

The evaluation criteria used for the selection of the BPEO and BPM are as follows:

1. Cost effectiveness

• life cycle cost of waste.

2 Operational feasibility

• existing or new technology

• international best practice

• regulatory constraints and challenges

• ease of operation.

3. Environmental and social acceptability

• public safety impact

• perceived risk and social acceptability

• environmental impact

• continual improvement potential.

4. Safety

• worker safety impact

• public safety impact

• accident risk

• safety impact reduction potential.

The aim of a BPEO study is to ensure that the reasoning behind a strategic decision, involving technical, scientific and more qualitative judgments (including their consistency with the overriding principles of precautionary action and sustainable development), is made visible.

The process to determine the BPEO for each waste or material category is presented schematically in Fig. 20.13 . In instances where management options for certain waste categories exist and those processes are author­ized or in the process of authorization, then that will be regarded as the preferred option (BPEO) for that waste category. It should be noted, however, that the authorization process includes evaluation of options and justification of selected actions.

This process entails the following in sequential order:

1. Identification of the waste or material. This includes radioactive waste or redundant radioactively contaminated materials/equipment for which no further need is foreseen as identified by the generator or waste operator.

2. Categorization/grouping of waste/material according to its possible management option. The categorization allows for grouping together of waste streams or material with the same attributes and that will be managed in the same manner.

3. Following the flow diagram for each waste/material category to find the BPEO and associated BPM.

• Recovery of uranium (source material). Some radioactive waste at Necsa contains uranium that could be recovered for re-use. Due to the high value associated with uranium, especially uranium in its enriched form, it is worthwhile to identify radioactive waste streams with recoverable quantities of uranium. The recovery route will be regarded as the BPEO for that respective waste stream. Different option evaluations will be performed in order to find the BPM for each waste stream in this waste category where recovery processes have to be developed.

• Waste/material clearable. Material or waste that conforms to the criteria for clearance will not be further treated but will be processed to demonstrate compliance with clearance criteria. The clearance route will be regarded as the BPEO for that category. The ‘clearance’ of radioactive material allows for the release of the material from nuclear regulatory control in terms of the compliance with clearance

levels. If the waste/material is clearable with further treatment, then it will be treated as such. If a treatment process exists, then treatment will be performed. If not, an exercise will be performed to find the BPM for treatment of that category as described previously. Existing treatment processes include chemical decontamination and smelting.

• Waste/material acceptable for re-use or recycling. This step allows for the release of material that does not conform to clearance criteria

but conforms to the criteria for re-use or recycling. If treatment is required to enable release of the material, the same procedure will be followed as described previously to determine the BPM.

• Waste/material acceptable for authorized storage.

• Waste/material acceptable for authorized disposal. This step allows for the release of material that does not conform to clearance criteria or for re-use or recycling but does conform to the criteria for author­ized disposal. If treatment is required to enable authorized disposal of the material, the same procedure will be followed as described previously to determine the BPM.

• BPEO waste management option study (regulated storage or regu­lated disposal). Waste that does not conform to any of the above categories requires an option study to find the BPEO for that waste category. The option study will determine whether the BPEO for a specific waste category will be regulated disposal or regulated storage. Once the BPEO for that waste category has been selected, the waste will be treated in accordance with existing processes or, if such processes do not exist, an option study will be performed to select the BPM.

There are three options when treating radioactive waste, namely volume reduction, removal of radionuclides from the waste and the change of the physical form and/or chemical composition. The nature and composition of the respective waste categories allow in most instances a decision to be taken on the BPM for that waste category. For example, in the case of the compressible waste type, the BPEO would be volume reduction. Different methodologies can, however, be applied in order to reduce the volume of the waste such as com­paction, incineration and segregation. Option studies are subse­quently performed in order to determine which of these methodologies would be the preferred option or for the respective waste category.

4. When the BPM evaluation process indicates the establishment of new facilities or processes or the significant modification of existing facilities or processes, then external regulatory approval is required. This will mainly entail the Environmental Impact Assessment (EIA) process as prescribed by the Environmental Management Act, No. 199 in order to obtain a Record of Decision (RoD) as well as the NNR’s nuclear authorization process.

To keep pace with the development of nuclear technology applications, temporary storage facilities have been constructed in China since the 1960s. The Notification on Strengthening Radioactive Environment Management

Arrangement was issued in the Temporary Regulations on Construction of Urban Radioactive Waste Repository in 1984. The Methods on Urban Radio­active Waste Management was issued in 1987 [12,13]. Temporary waste storage facilities are constructed on a provincial basis. Each province (or autonomous region, or municipality directly under central government) builds one such facility to accommodate wastes arising from research, teaching, medicine and other applications of radioisotope and nuclear tech­nology within the province. Provincial environmental protection agencies have set up special organizations staffed with specialists responsible for supervision and environmental monitoring. The Criteria on Siting, Design and Construction of Application Waste Storage Facility was issued in 2004 [14] and requires the modification and extension to be carried out for exist­ing storage facilities to meet the new requirements. At present, special funds have been appropriated for this purpose. It also requires an environmental impact assessment to be made prior to such modification and extension, which cannot be implemented without approval by the relevant agencies. By the end of 2010, a total of 31 waste storage facilities, together with one centralized storage facility for spent radioactive sources, had been con­structed and/or upgraded in compliance with the new requirements. At the end of 2006, these facilities had received 64,572 m3 of disused sealed sources, of which 49,741 m3 are in the provincial storage facilities, and the remainder is in the national centralized facility.

Radioactive materials have been released to the environment from the damaged cores of units 1, 2, and 3. Grambow and Poinssot [6] considered the extent of core damage, suggesting from the available evidence that not only the UO2 in the fuel but also the zircaloy cladding and steel melted forming a quenched melt (also seen at Three Mile Island and Chernobyl) called corium. They believe substantial amounts of volatile fission products (FPs), such as Cs and I, were released during melting, but that the less vola­tile FPs and actinides were incorporated into the corium. This corium thus contains most of the most radiotoxic species and presents a very large and long-term challenge. The escape of radionuclides from the damaged cores occurred both through atmospheric release and through water leaks to the sea. It is considered that the major part of the atmospheric release was by the unexpected leak of the containment vessel, whilst deliberate venting through the suppression chamber to reduce gaseous pressure gave rather limited result. A controlled release of very low-level contaminated water from the central waste treatment facility (ca. 4 Bq/cm3) to the sea was done once, but it has little significance in terms of the total amount of contamina­tion. The major release to the sea was by (1) fall-out of the atmospheric release to the sea, (2) carry over by rainwater, and (3) leakage of highly contaminated water via underground cable (sub-drain) pits.

The extent of radionuclide release to the atmosphere has been evaluated several times by government officials and Tokyo Electric Power Company (TEPCO), but it is still very much an estimate [5-13] . This can be seen in Table 24.2 which compares the estimated release of cesium-134, 137 and iodine-131 in March 2011 by several estimation studies. The differences in the data arise from the different assumptions used about the status and progress of the accident, due to unreliability in instrumentation data as well as in the monitoring data under the severe conditions. The estimate by TEPCO, which showed larger values than government authorities, was obtained by comparing the result of dose monitoring as well as of the analysis of radioactivity deposition onto surfaces, and the results of envi­ronmental dispersion calculation code. The calculation is based on an assumed rate of evolution of radioactive particles, and thus still has some uncertainty.

Cesium-134, 137 and iodine-131 are the nuclides that mostly decide the radiological impact on the public, among which iodine-131 due to its short

JAEA: Japan Atomic Energy Agency; NISA: Nuclear and Industrial Safety Agency, Japan; NSC: Nuclear Safety Commission of Japan; IRSN: Institut de radioprotec­tion et de surete nucleaire.

* INES index: Radiologically equivalent value to iodine-131. The International Nuclear and Radiological Event Scale (INES) was developed in 1990 by the IAEA and the OECD Nuclear Energy Agency (OECD/NEA) with the aim of communicat­ing the safety significance of events at nuclear installations [ 14] and Section 1.4.6). If there is an atmospheric release from a nuclear facility then a radiological equivalence to iodine-131 is calculated using conversion factors. For example, the actual activity of the isotope released should be multiplied by some factors (given in [ 14]) and then compared with the values given in the definition of each level. An event resulting in an environmental release corresponding to a quantity of radioactivity radiologically equivalent to more than several tens of thousands of TBq of iodine-131 is rated to the highest level 7 according to the INES scale.

half-life is only significant for a couple of months after its release. The esti­mated amount of the other released radioactive nuclides can be seen in Table 24.3, which was evaluated in June 2011 and corrected in October 2011 by the Japan Nuclear and Industrial Safety Agency (NISA). In a general sense, the nuclides other than cesium-134, 137 and iodine-131 have less radiological impact because of their lower radioactivity, low specific radio­logical effect, as well as short half-lives. It should be noted that the amount of strontium-90 believed to have been released is about 1/100 of that of the sum of cesium-134 and 137, and that of plutonium-239 is about 10-7 of that.

The amount of radioactivity released to the sea via the sub-drain pit was estimated to be 4.7 PBq in total of cesium-134, 137 and iodine-131 [8], which is much less than the atmospheric release. Due to the leak of the water from the sub-drain pit, the radioactivity concentration in the seawater sampled at the discharging point showed as high as 105 Bq/L of cesium-137 at the

released radioactivity spread in other directions, such as to the south. The dispersion of radioactive materials was tracked by the Preparatory Com­mission for Comprehensive Nuclear-Test-Ban-Treaty Organization (CTBTO) [8] which is a monitoring system designed to detect nuclear explosions. CTBTO reported the large-scale radiation leak resulting in the 20 km exclusion zone being set up around the power plant and people within the 20-30 km zone being advised to stay indoors.

A total of 68 underground detonations were conducted in tunnels con­structed in unsaturated zeolitized volcanic rocks of Rainier Mesa and Shos­hone Mountain (tunnel beds); all were located well above the regional groundwater table. Two detonations were in vertical shafts near the water table. The Rainier Mesa and Shoshone Mountain sites form plateau high­lands that demarcate the approximate eastern edge of the thick accumula­tions of volcanic rocks formed within the Amargosa Desert rift zone (Figs

26.3 and 26.6). The migration pathway of radionuclides released during testing beneath Rainier Mesa is generally downward through the unsatu­rated zone, complicated by local zones of perched water; travel time to the regional water table may be substantial. Additionally, there are local losses of radionuclides from drainage into tunnels constructed to host the

26.5 Generalized geologic map of the Rainier Mesa-Shoshone Mountain (RM-SM) corrective action unit showing the domain area for numerical models of groundwater flow and radionuclide transport at sites of underground testing. White = Quaternary/Tertiary alluvium; light gray = Miocene volcanic rocks; cross-hatch = Quaternary/Pliocene basaltic rocks; diagonal line = Mesozoic granitic rocks; dark gray = Precambrian and Paleozoic sedimentary rocks. Dashed line is the RM-SM hydrostratigraphic framework model boundary. Solid line is Nevada National Security Site boundary. Double-dash line is caldera structural margin (buried). Dots show the location of 68 underground detonations in the Rainier Mesa-Shoshone Mountain corrective action unit (as well as those in western Yucca Flat and eastern Pahute Mesa).

underground tests and from discharges from the tunnels into unlined drain­age ponds.

Groundwater beneath Rainier Mesa and Shoshone Mountain is recharged primarily by infiltration through the thick unsaturated zone beneath vol­canic highlands; downward infiltration through the zeolitized tunnel beds is locally aided by flow in zones of discontinuous fractures forming local perched water (Russell et al, 1987). The amount of underflow beneath the mesa from regional groundwater flow of the DVRFS is poorly constrained and the recharge mound beneath the mesa highlands may not be well con­nected to the regional groundwater flow system because of local juxtaposi­tion of clastic sedimentary confining units associated with thrust faults at the eastern edge of Rainier Mesa (Fenelon et al., 2008).

Liquid LLW and ILW wastes are either evaporated or mixed in cement, while solid wastes are crushed, incinerated, compacted or cemented

beforehand. All waste types are packaged in standardized and approved containers after processing. Cylindrical concrete containers are generally used for solidified waste, while unconditioned waste is sealed in iron waste containers with welded lids. LLW and ILW currently in interim storage will be disposed of in the Konrad facility. Existing waste at Morsleben will be considered as geologically disposed upon completion of the facility closure. No final decision has been made as to the ultimate disposition of waste currently stored at the Asse facility.

D. JACKS ON, A. BAKER, R. GEORGE and S. MOB BS, Eden Nuclear and Environment Ltd, UK

DOI: 10.1533/9780857097446.2.509

Abstract: The United Kingdom has a long history of nuclear development. Waste management principles and strategies have evolved over this period, together with technical developments allowing modified waste management regimes. High and intermediate level wastes reflect both current arisings and legacy wastes, with disposal and storage options being explored and implemented. In recent years, low level waste strategy has been further clarified to expand options available for safe and cost-effective disposal. Challenges with respect to contaminated land and delicensing of decommissioning sites are recognised.

Devolution of waste management responsibilities within the UK is leading to some divergence in national policies, particularly with respect to higher activity wastes.

Key words: Nuclear Decommissioning Agency (NDA), Magnox, advanced gas-cooled reactor (AGR), reprocessing, waste policy.

16.1 Introduction

The United Kingdom has a long history of nuclear development, which, for convenience, can be traced from the post-war weapons programme and, later, the civil use of nuclear power. Research and production sites in England at Harwell (Oxfordshire), Sellafield (Cumbria), Springfields (Lan­cashire) and Capenhurst (Cheshire) were established in the 1940s and, in Scotland, the Dounreay site (Caithness) followed in 1954, initially to develop the fast breeder reactor.

The UK’s first commercial nuclear power reactor began operating in 1956 and, at its peak in 1997, 26% of the nation’s electricity was generated from nuclear power. Nuclear reprocessing facilities were also built to deal with the increasing demand from both military and civil programmes. Since then a number of stations have been closed, and others are scheduled to follow over the next decade. Of the currently operating stations, lifetime exten­sions may be granted for some sites, allowing for continued generation until

replacement generating sources become available. Locations of all major nuclear licensed sites in England and Wales are presented in Fig. 16.1.

This account of radioactive waste (RAW) management in England and Wales is oriented towards the strategic and environmental issues arising from the management of RAW from the nuclear industry. It also addresses the structure of the nuclear industry and the sources, types and classification of RAW.

Approximately one million m3 of solid RAW has been disposed of in the UK to date (NDA and DECC, 2011). Current wastes identified, plus pro­jected wastes over the next century or so, amount to around 4.7 million m3 in the UK. About 97% (4.6 million m3) of the total volume of RAW antici­pated has already been produced. Some has been processed, and is being held in stores, but most is contained within existing nuclear facilities, includ­ing reprocessing plants and nuclear reactors, and will not be processed until these are shut down and dismantled. This waste is the legacy of past and current civil and military nuclear programmes. About 3% (150,000 m3 ) of the radioactive waste total has yet to be produced. This waste is that forecast from the future planned operations of the existing nuclear power industry, from ongoing defence programmes and from the continued use of radioac­tivity for medical and industrial purposes.

Current and projected radioactive waste volumes for England and Wales are summarised in Table 16.1 (NDA and DECC, 2011).

The Committee on Radioactive Waste Management (CoRWM) produced its recommendations to the UK government and devolved administrations on the long-term management of HAW in November 2006 (CoRWM, 2006).

At that time, as a sponsor of CoRWM, the Scottish government was content with CoRWM’s recommendations. The main recommendation was that geological disposal was the best available approach for the long-term man­agement of HAW. One of the qualifying conditions was the need for robust interim storage until geological disposal could be implemented, which could take up to 40 years.

The Scottish government, however, changed its view in June 2007 when Richard Lochhead, Cabinet Secretary for Rural Affairs and the Environ­ment, announced that the Scottish government ’s policy for the long-term management of HAW was to support long-term, near-surface, near-site storage so that waste is monitorable and retrievable and the need for trans­porting the waste is minimal (Scottish Government, 2010a).

Consequently, the framework for implementing geological disposal, Man­aging Radioactive Waste Safely (MRWS), published as a White Paper in June 2008 (UK Government, 2008) was not sponsored by the Scottish gov­ernment. The White Paper noted that the Scottish government supported long-term interim storage and a programme of research and development.

Thereafter, the Scottish government embarked on a process (CoRWM, 2011) to develop a more detailed statement of its own HAW policy which included significant stakeholder engagement. From January to May 2010 the Scottish government consulted with the public and stakeholders on a draft Detailed Statement of Policy for Scotland’s HAW (Scottish Govern­ment, 2010a) and supporting documents which comprised a Supplementary Information report (Scottish Government, 2010b) and an Environmental report (Scottish Government, 2010c).

The Scottish government published its Policy (2011a), Summary of Com­ments (2011b) and Post Adoption Strategic Environmental Assessment Statement (2011c) on 20 January 2011. An additional seven reports pro­vided supporting information.

The policy sets out in detail the Scottish government’s position on a wide range of HAW issues. In summary the key points are: [29]

• When disposal is employed, the facilities should be near surface (no more than a few tens of metres below surface) and near to the site where the HAW is generated.

• Disposed HAW must be able to be monitored and retrievable.

• The policy will be reviewed every ten years to assess whether technolo­gies have developed sufficiently to warrant a change to the policy.

The policy for HAW in Scotland now differs from the rest of the UK. The Scottish government has rejected at present the concept of deep geological disposal for HAW that cannot be disposed of in near-surface facilities. The Scottish government is not taking part in the MRWS process to identify a site for, and develop, a geological disposal facility (GDF). The policy does not contain the concept of volunteerism by communities for HAW facilities which is fundamental in the MRWS process.

Operators with HAW at facilities in Scotland which is unsuitable for near-surface disposal and who previously had planned for disposal of that HAW in the projected UK GDF now need to plan for new stores and longer storage periods in the absence of an identifiable final end-point.

The generally accepted understanding of disposal is that there is no inten­tion to retrieve the waste. In practice, this means specific retrieval features are not included in the disposal facility design. Also monitoring is applied to the surrounding environment rather than the waste itself. In the Scottish context, while there may be no intention to retrieve, the policy requires that HAW in near-surface facilities must be able to be monitored and retriev­able. This introduces additional requirements for designers and operators of Scottish HAW disposal facilities for monitoring of the waste itself in the facility and for including features to enable retrieval of the waste.