In 1971 the central radiation safety organization, Radon, looked at the total of its results over the first ten years of operation. The enterprise already had a well-developed infrastructure, skilled staff, methods and technologies for RAW localization and a substantial amount of operational experience. Hot cells equipped with manipulators were put into operation and were allowed to operate remotely in RAW and SIS procedures. In the same year, the director of Radon (I. A. Sobolev) prepared a doctoral thesis dedicated to the problems of RAW cementation, which he successfully presented to the scientific and technical council of the Institute of Inorganic Materials under the auspices of A. A. Bochvar. To allow the routes used by the special RAW transport to be monitored for radiation levels, two mobile radiomet­ric laboratories were set up. The activity of Radon then became known abroad: the first technical visits by foreign representatives of the atomic

industrial forum of Japan, parliamentary groups from Sweden and scientists from Canada took place in 1975.

From 1976 to 1980 new facilities were installed at Radon for RAW bitu — minization (URB-200), combustion (USGO-80) and compaction (BA — 1330). During the same period, a facility for water purification (EDU-500) was assembled, and test runs were carried out. Another considerable achievement was the implementation of a pilot industrial facility for RAW combustion or thermal treatment: RAW combustion was carried out for the first time in the history of Soviet industrial tests in NPPs. The results of these tests formed the basis for the design and creation of the technological complex for the Kursk NPP. The application of a ‘dry’ system for the treat­ment of off-gases helped to solve the problem of secondary LRAW genera­tion. To obtain bitumen compounds, the UBD-200 facility was developed, with elements of mechanization and automation for the whole process from RAW loading to the unloading of the end product (i. e., the bitumen compound).

Radon had swiftly become a multi-purpose scientific and technological complex. The new technological developments introduced at Radon were used at other facilities for RAW processing and disposal. In 1977 the State Committee on Inventions and Discoveries awarded two certificates of authorship to the staff of Radon for patented inventions. Radon’ s main achievement in 1978 was the establishment of the new high-capacity facility for LRAW bituminization (URB-8). Work was also carried out on the selec­tion of the correct composition for LRAW vitrification. Under the manage­ment of the corresponding member of the Russian Academy of Sciences and the director of the Institute of Inorganic Materials (A. S. Nikiforov), the process of high-level LRAW vitrification at hot cells was investigated, using simulators of high level LRAW from Mayak. In 1979 the containment facil­ity for HLW and SIS was completed. Research into hot cells was carried out, including studies into the content of intermediate and high level waste. Experiments into waste vitrification were continued, alongside new studies into plasma methods, and a new facility for RAW combustion, known as Fakel, was completed.

In 1980, the Council of Ministers of the USSR converted the enterprise into the Moscow Science and Production Association ‘Radon’ (MosNPO Radon); after that, the organization became responsible for the methodical and scientific management of RAW at special nuclear facilities across the USSR. In 1984, before the Chernobyl disaster, the government of the USSR charged MosNPO Radon with the organization and implementation of detailed radiometric monitoring of Moscow city and the Moscow region in order to define areas with possible radioactive contamination. Furthermore, in 1986, MosNPO Radon, Minsredmash and the USSR Academy of Sci­ences were jointly entrusted with scientific research and experimental design with the aim of improving RAW processing technologies such as combustion, vitrification, cementation, etc.

The need for protective actions became evident very soon after the Cher­nobyl accident occurred. Activities for area decontamination were part of an extensive set of short — and long-term environmental countermeasures, applied to protect workers and the public from radiation. These counter­measures involved large amounts of human, economic and scientific resources. According to IAEA (2001), such countermeasures included:

• reduction of radionuclides release from the destroyed reactor (in the early stages of the accident),

• evacuation of population and its resettlement,

• construction of the ‘Shelter’ object (SO),

• decontamination of the soil, buildings and installations,

• disposal of the RAW resulting from the decontamination measures,

• surface and groundwater protection,

• restriction of access to the contaminated areas and the prohibition of economic activity,

• changing the type of forestry and agricultural activities,

• ban or limitation of the consumption of contaminated foodstuffs,

• reduction of radioactive contamination of agricultural products,

• information to the population, social and other supplementary measures.

Large-scale decontamination and clean-up activities were performed between 1986 and 1989 both within the Chernobyl Exclusion Zone, and in the cities and villages of the USSR most contaminated after the Chernobyl accident. This activity was performed usually by military personnel and included removal and disposal of contaminated soil and civil constructions, cutting and disposal of contaminated forests, washing of buildings with water or special solutions, cleaning and washing of roads, and decontamina­tion of open water supplies. The decontamination has produced a huge amount of radioactive waste, which was collected in numerous RWDP and RWTSP and has created a problem for its final disposal. More than 800 waste localization sites were created within ChEZ and 47 outside of it, with a total volume of more then 106m3 . The reliability of these sites is a cause for concern, and the problem needs to be solved in the future. The greatest amount of RAW arose during decontamination of the site of the destroyed unit 4 in the course of the ‘Shelter’ object (SO) construction. The object is classified now as a temporary storage of radioactive waste (RSNU, 1997). The SO, RWDP and RDTSP are described in Section 11.4.3.

The efficiency of various measures for protecting workers and public from radiation was assessed in IAEA (2006) . According to this source, depending on the decontamination technologies used, the dose rate was reduced by a factor of 1.5-15. However, the high cost of these activities hindered their comprehensive application to all contaminated areas. Due to these limitations, the actual effectiveness of the decrease in annual exter­nal dose was 10-20% for the average population and ranged from about 30% for children to less than 10% for outdoor workers.

It should be noted that in many cases the decisions about decontamina­tion in 1986-1989 were of a political nature. In many cases such political decisions were in contradiction with conclusions reached by cost-benefit analysis. Therefore, the resulting effect in reducing the exposure dose was achieved by unnecessarily high costs in human and economic resources.

The NPP operational radioactive waste category incorporates the waste generated in both the operation and decommissioning of NPPs. RAW gen­erated in institutional applications are also included in this group and in some countries are even managed (entirely or only for some steps) together with NPP waste. This category is the biggest from the point of view of the waste volume generated and covers a wide spectrum of wastes with respect to their radiological, chemical, physical, aggregate and other properties. Therefore the classification system is complicated, usually reflecting coun­try-specific approaches, national waste management strategy, regulations, and often technological and other aspects. The IAEA in its Radioactive Waste Management Status and Trends [ 1] has recognized more than 40 different country-specific RAW classification systems. It is obvious that under such conditions it is difficult to unify approaches, evaluate and compare national systems, and make a plan for optimization and recom­mendations for safe RAW management.

The IAEA has made a systematic effort to develop a unified and inter­nationally accepted classification (categorization) of RAW for a long time in order to facilitate communication between involved parties in various countries and organizations. At the beginning of the 1990s the first comprehensive IAEA document was published on Classification of Radioactive Waste [2], which based the RAW classification system on aspects related to waste disposal safety and proposed the boundaries of individual RAW categories using International Commission on Radiological Protec­tion (ICRP) recommendations for annual individual and collective doses for public. The two main radiological parameters used for the RAW classifica­tion system are the radioactive material concentration and the half-life of critical radionuclides. This proposed system has been accepted for develop­ment of a national RAW classification system in many countries.

Further evolution resulted in a new IAEA RAW classification system published in 2009 [3] [ which is more generic and almost exclusively based on long-term waste disposal safety considerations (in other words, for each RAW class there is a specific proposed/assigned waste disposal route). This recent waste classification system was modified in order to reflect a better relationship between RAW categories and the safety aspects of the consid­ered disposal options. While the 1994 IAEA document provided basic numerical boundary values for various RAW classes, the 2009 document offers only a general approach and leaves the development of more exact figures to the individual national regulations. For illustration and a better understanding of the boundary parameters and values for distinction of the

RAW categories, the data from the 1994 IAEA document [2] are given in Table 2.1. In spite of the fact that the 1994 IAEA document [2] is no longer applicable, while the updated IAEA 2009 document is available [3], these figures are still used as informative values in preparation of the national classification systems.

In accordance with the present IAEA approach to RAW categorization [3], six classes of waste are proposed for consideration in preparation of the national classification systems. It should be noted that precise values of applicable total activity content and eventually activities (activity concen­trations) of the most significant individual radionuclides for each waste category shall be specified on the basis of safety assessments for individual

Table 2.1 Typical characteristics of waste classes [2]

Waste classes Typical characteristics Disposal options

a IAEA draft document: IAEA Clearance levels for Radionuclides in Solid Materials: Application of Exemption Principles, Safety Series No. 111-G-1.5, in preparation in 1994, IAEA, Vienna. Later reconsidered, rewritten and published as Reference [4].

disposal route and disposal site. This is the responsibility of the national authorities and the values used can differ significantly in individual countries.

The international safety standards for RAW management are grouped into those relating to pre-disposal management of RAW [39], disposal of RAW [40] , decommissioning of facilities [41] and remediation of contaminated buildings and areas [42]. These are elaborated in the sections below. The safety requirements set out in the standards are aimed at governments, regulators and operator organisations carrying out waste management activities and those parties responsible for contaminated environments.

At all nuclear power reactors, irradiated fuel is first transferred from the reactor into water-filled storage pools. The fuel is stored in geometric arrangements that eliminate the possibility of an inadvertent criticality. Over the years, as irradiated fuel has continued to accumulate, utility opera­tors have implemented storage technologies that have increased the capac­ity of the storage pools, without increasing the size of the pools themselves. This has been achieved by re-racking the fuel bundles into storage arrays that incorporate neutron absorbers. Criticality control through adding soluble boron to the pool and through placement of neutron-absorbing storage rack inserts have further increased pool storage capacity. These technologies have made it possible to space the fuel in the storage pool at nearly the same packing density that is achieved in the reactors themselves (Kessler, 2010). Figure 5.1 shows a typical pool for storage of irradiated fuel.

Despite increases in pool storage capacity, utilities are continuing to move towards adding dry storage capacity. This need is driven principally by continuing delays in the establishment of geological repositories, but the experience with pool-stored fuel at the Fukushima Daiichi nuclear complex following the earthquake/tsunami event might also provide impetus to more quickly move irradiated fuel to dry storage. Although early develop­ment of dry storage technologies was directed simply at fuel storage, the approach has evolved into developing containers that serve both as a storage container and as a shipping cask, so-called dual — or multi-purpose dry storage systems. The advantage of this approach is that the fuel, once removed from the storage pool, has to be handled only once; since the storage container and shipping cask are one and the same, there is no need to move the fuel from the storage container to a suitable shipping cask. A

number of dual-purpose dry storage containers have been developed and these have been well described elsewhere (Kessler, 2010). Figure 5.2 illus­trates a typical dry fuel storage facility.

Composites can be thought of as multi-barrier waste forms. Usually a com­posite waste form is required to meet a specific waste form criterion, e. g. heat loading, respirable fines, compressive strength, etc. A single-phase or multiphase crystalline ceramic or even a glass can further be encapsulated in a metal, a glass, or an encapsulant waste form such as cement, geopoly­mers, hydroceramics, bitumen, etc. The encapsulant phase offers a second level of protection to the release of radionuclides or hazardous components in the waste form as shown in Table 6.10. Composites include many GCMs such as glass bonded sodalites that have already been discussed (on page 214). Composites can also include deteriorated cement waste forms that are remediated by encapsulation (see Table 6.10). A few examples are given below and others are shown in Fig. 6.3.

Metal matrix

In metal matrix waste forms a metal is used as the encapsulant for either glass or crystalline materials in which the radionuclides or waste species are already atomically bonded. The advantages of this type of encapsulation include (1) improved thermal conductivity of the waste package, (2) poten­tially decreased leach rates of radionuclides because of the metal matrix encapsulation, (3) improved mechanical strength and decreased dispersa — bility on impact, and (4) improved radiation protection during handling [152, 153]. The encapsulation of waste forms in metal matrices was pursued in the US and developed full scale in PAMELA, which was a joint Belgian — German project located in Belgium.

Vitromelt is a composite waste form in which glass beads (0.5 cm) are embedded in a metal matrix (usually a Pb alloy) [154-156] . For example, waste immobilized in calcium silicate pellets was encapsulated in a lead matrix. In one variation of the commercial PAMELA vitromelt process, phosphate glass beads containing HLW were produced by passing molten glass through nozzles. The beads were subsequently fed into a container and infiltrated with molten lead alloy to produce a composite waste form (‘vit — romet’). The beads, with a diameter of 0.5 cm, occupy up to 66% of the total volume. Increased thermal conductivity of the waste form leading to lower waste temperatures is one of the most important advantages of this product.

I n studies related to vitromelts, immobilized waste pellets have been coated with pyrolytic graphite, before encapsulating in a metal matrix, in order to improve the leach resistance. Application of other coatings has also been reported, including alumina, titania, silica, silicon carbide, chromium silicide, and chromium oxide, together with a variety of metals including Ni, Fe, and Mo. Dual coatings of pyrolytic graphite and alumina have also been reported. Metal matrices have included Pb-based alloys (e. g., Pb-Sb, Al, Sn), Al-based alloys (e. g., Al-Si, Cu, Ti), and Cu. Particles can be coated by conventional ceramics (e. g., Al2O3, TiO2 . or SiO2) or by carbon products (e. g., PyC, Cr7C3 . or SiC), glass (borosilicate or aluminosilicate), or metals (e. g., Ni, Si, or Fe) before being encapsulated). Uncoated, sintered super — calcine pellets have been encapsulated in vacuum-cast Al-12Si and glass-coated, sintered supercalcine pellets encapsulated in vacuum-cast Al-12Si. Supercalcine pellets have also been PyC/Al2O3 coated before encapsulation in gravity-sintered Cu.

Cermets are related composite waste forms in which radionuclides in the form of small oxide or silicate particles +1 mm in size are dispersed in a metal matrix [152]. The unique aspects of the waste form are the very fine scale on which the radionuclide-containing phases are dispersed, the fact that the alloy is primarily composed of hydrogen reducible metals which are already in the waste, the high thermal conductivity, and reduced leach rates due to the alloy encapsulation. Developmental work on cermet was performed using simulated wastes, radionuclide-containing simulated wastes, West Valley acid THOREX wastes, and SRS HLW sludge and un­neutralized SRS wastes. Waste loadings of up to around 30% have been reported. The addition of elements in excess of stoichiometric requirements is used to guarantee the formation of specific ceramic phases, e. g. excess Al and Si to ensure the formation of pollucite.

A central feature of the ER process is characterization. In this context, characterization refers to those investigations, specifically including meas­urements, undertaken to provide information and data about the contami­nation and affected site environment. Characterization steps usually taken include: [23]

• design of the selected remediation option;

• implementation of the remediation option; and

• verification and/or monitoring of the remediation.

Characterization is a necessary prerequisite to provide critical information and data for each assessment step in this process. Multiple characterization activities are common, with each characterization activity focused on gath­ering the information essential for the particular type of assessment being conducted.

While the general process of dealing with a potentially contaminated site is applicable to most problems, it may result in a range of characterization activities that vary widely in terms of scope, cost and schedule. For example, a small ‘hot spot’ of radioactively contaminated soil resulting from a recent small spill may be surveyed, hand shovelled up into a small container for proper disposal elsewhere, and the soil replaced with clean soil in a few hours. The related characterization activities would have amounted to field survey instrument measurements of radiation prior to and after the hand shovelling.

Alternatively, the source of contamination may have been a leak of radioactive material that contaminated not only the surface soil in the immediate vicinity of the leak but also distant areas, the subsurface soils and groundwater. Migration of the contaminant might now threaten the environment and population away from the leaking source. In this instance, the components of the assessment process may be more complex and, con­sequently, the characterization activities may be more in number, more elaborate, and require years to complete.

Major factors to be taken into account in site characterization include:

• Characterization can be a large consumer of project resources. Mistak­enly, its practical importance to solving the problem may not always be understood or appreciated. In some instances, the characterizations may be the ‘last word’ measurements (e. g., for peripheral areas) and, as such, their credibility is vital.

• The amount of characterization should be proportionate to the extent of the likely remediation effort. Over-characterization can result in a disproportionate fraction of the budget being spent on meas­urements, leaving insufficient means to carry out acceptable remediation.

• Characterization should be adequate to allow a properly designed reme­diation; one that does not involve excessive amounts of unnecessary effort or environmental damage.

• Characterization efforts should be sufficient to demonstrate the exist­ence of clean areas and to provide credible assurances that un-remedi­ated areas are safe.

• Characterizations should have a sufficiently broad focus that any other unknown contaminants are detected at a stage when they can be dealt with efficiently.

• The characterization, in the first instance, and the subsequent remedia­tion should not make things worse by ill-advised first attempts that magnify or spread the problem. A guiding principle can be ‘first, do no harm’.

The reader should note that all factors listed above are relevant to the subsequent generation and management of wastes.

Details on characterization methodologies, techniques and instruments can be found in IAEA (1998) and ITRC (2006). Practical experience, including also R&D work, can be found in IAEA (2000). It should be noted that characterization plays an essential role in the end of an ER project to certify compliance with end-state criteria and allow the planned reuse of the site. Details on post-ER characterization are given in IAEA (1999a) . Figure 8.2 shows post-decontamination measurements of soil by Radon company (Russia).

The improvement and development of the state system for the accounting and control of radioactive material (RAM) helps to prevent the unlawful use of such materials and to facilitate radiation and environmental safety [ 27]. The national system of state accounting and control of RAW and RAM (NSSRMWAC) was created in accordance with Federal Law No. 170-FZ ‘Nuclear Power Use’ in November 1995. The purpose of the system is to register the existing quantities of RAM and RAW in their storage and disposal locations, to prevent loss, unauthorized use or theft of RAM and RAW, and to provide the national authorities and nuclear safety regulators with information regarding the availability, movements, exports and imports of RAW and RAM.

As per the Russian Government Decree No. 1298 of October 1997 on the Approval of Rules for the Organization of a National System of Accounting for Radioactive Materials and Wastes, the Russian Ministry of

(d)	(e)

10.9 Cementation by ash residue impregnation in 200 l barrel:

(a) experimental-industrial installation, (b) initial radioactive ashes, (c, d) container and probe, (e) end product (cement compound with filling of the ash residue to 70% by mass).

Atomic Energy (Minatom, later Rosatom) was appointed responsible for all activities associated with the functioning of the accounting system for RAM and RAW. During 1998-2000, Minatom prepared the key legal acts and procedural documents to ensure the successful creation and operation of this system, including:

• the Statute of State Accounting and Control of RAM and RAW in the Russian Federation,

• procedural recommendations for primary inventory checks of RAW and RAM,

• forms for the provision of accounting data relating to RAW and RAM, along with instructions for the completion of these forms,

• forms for federal statistic supervision 2-TP (for radioactive materials), 2 TP (for radioactivity) along with recommendations for completion.

This documentation allowed a primary inventory of RAW and RAM to be introduced on 1 July 2000, with information starting to be collected from sites and analysed since 1 January 2001. Functionally, the organization of the national system of RAW and RAM accounting and control (Fig. 10.10) includes three levels of authority: federal (Rosatom), regional (regional executive authorities) and departmental (federal executive authorities). Under this system, Rosatom acts both as a federal and a departmental authority.

10.10 Organizational functional diagram of NSSRMWAC.

The federal level of the accounting system shall provide:

• federal-level accounting of RAW and RAM;

• collection and analysis of RAW and RAM accounting information at the regional and departmental levels;

• formation of databases for the state cadastre of RAW, RAW storage and disposal sites, and contaminated territories that are within the area of responsibility of the supervising organization;

• organization of information exchange between the authorities that control the accounting system at the federal, regional and departmental levels;

• development of scientific, procedural and technical developments that help create, operate and enhance the accounting system, providing the results to organizations involved with RAW and RAM control account­ing at all levels;

• development of regulations and code documents (standard accounting forms for radioactive materials and waste, quantity and radionuclide composition measurement procedures, etc.), and, together with other federal executive stakeholders, of inter-compatible software for the databases;

• co-ordination of federal-level efforts;

• information for the federal authorities and nuclear regulators and other stakeholders regarding the availability, movement, imports and exports of radioactive materials and waste as required for these bodies to exer­cise their authority;

• management of the Information and Analytic Centre of the National System of Accounting and Control of Radioactive Materials and Radio­active Waste, providing information and analysis that helps the system to function at its federal level;

• co-operation with foreign nations on issues regulated by international agreements and programmes (projects) related to accounting of RAW and RAM.

The regional and departmental levels of the system perform similar func­tions scaled down to their areas of responsibility. Overall supervision of the system is entrusted to the Federal Service of the Ecological, Technical and Atomic Supervision of Russia (Rostechnadzor), which also licenses the corresponding forms of activity, and controls the observance of the estab­lished standards and rules for radioactive materials and waste management.

The system handles three flows of information. Information is supplied by every site, organization or subsidiary within 10 days of any operation involving the movement or change of status of RAM and RAW at its present location, transfer to other sites or legal persons, or receipt of RAM and RAW. This ensures that radioactive substances are tracked throughout the entire management process, from the time of generation through all movements between sites and enterprises up to their classification as RAW. Similarly, movements and transitions of RAW are also tracked through to their placement in long-term storage or disposal locations. As the move­ment of RAM and RAW requires that reports be submitted by both the transferring and the receiving party, the security of the RAM and RAW during transfers between legal bodies is ensured.

Annual reporting carried out by sites and enterprises using the forms supplied by the federal statistics services not only helps to monitor whether operations involving RAM and RAW are being reported correctly, but also further analyses additional information about radioactive releases and effluents, contaminated land, and so on. The periodic taking of inventories (performed annually for RAM and once every five years for RAW) helps in the generation of accurate data regarding the availability and character­istics of these materials at their storage locations, as well as collecting additional information about the characteristics of the storage locations.

The creation and operation of this system has vastly improved the accounting of RAM and RAW at Russian sites. Another important factor in this improvement was also played by the introduction of the federal regulation Main Rules of Accounting for Radioactive Materials and Waste in Organizations [28] in 2006; increased supervision by Rostechnadzor with regard to the observance of these rules also played a key role. The results of the analysis of the information stored in the national system of account­ing for radioactive materials and waste have been used for the production of a range of codes and guide documents, the most important of which is the Federal Program ‘Assurance of Nuclear and Radiation Safety for 2008 and until 2015’ and the Federal Law ‘On Radioactive Waste Management’.

According to Polish Atomic Law, RAW is classified according to its activity level or exposure measured at the surface as low-, intermediate — or high — level waste. These categories can be divided into subcategories taking into account their half-life or generated thermal power. Another category is used sealed radioactive sources which are divided into short — or long-lived low-, intermediate — or high-level.

High-level waste independent of the SF’s activity is mainly high uranium content waste especially from spent nuclear fuel or waste remaining after its reprocessing. Producers of such waste were research reactors. The first of these, Ewa, is already closed, while the second, Maria, is still working in the Institute of Atomic Energy (IEA) — POLATOM in Swierk near Warsaw. In Poland, spent nuclear fuel or uranium ore is not reprocessed, so there are no further sources of high-level waste. All other waste, generated by industry, hospitals, scientific and educational institutions, are characterized as low or intermediate.

For treatment, the radioactive waste in Poland is divided into gaseous, liquid and solid waste. The solid waste is divided into compressible or non-compressible.

The first Polish research reactor (Ewa), which was a Russian tank-type, was shut down in 1995 after 35 years of operation. The second research reactor (Maria), which is a 30 MW pool-type reactor, has been in operation in the Institute of Atomic Energy — POLATOM since 1975. The Maria reactor was designed mainly for material testing. Between 1985 and 1992 the reactor was shut down and modernised. The reactor is planned to work until 2020 and then, after further modernisation, from 2020 to 2050. The Maria reactor is now one of the best research reactors in Europe. It has power higher than 15 MW and neutron flux higher than 1 x 1014n/cm2, and is mostly used to produce radioisotopes, for materials testing, activation analysis, etc. Initially, the Maria reactor was supplied with highly enriched uranium (HEU) fuel with enrichment level up to 80%. Since 2002, for non­proliferation reasons, a low enriched uranium (LEU) fuel has been used with enrichment level up to 36%.

Defence-related wastes tend to be simpler than those from commercial nuclear applications. Wastes derived from Pu production contain high levels of sodium, due, for example, 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 fission products 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 1.5), and although there are very large compositional ranges for the constituents, it does highlight the lower proportion of fission products but higher proportion of actinides present in defence waste.

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 US and Russia to reduce their warhead stockpiles. Under the 1993 Non­Proliferation and Export Control Policy, the US declared 55 tons of pluto­nium surplus to national security needs. A similar quantity was also declared

surplus by Russia. These quantities may be further increased following the 2010 US-Russia strategic arms reduction agreement. It is planned to utilise this where possible in MOX fuel.

Finally, weapons testing has left a legacy of contaminated sites worldwide. These include Semipalatinsk and West Kazakhstan (Kazakhstan), Novaya Zemlya (Russia), Lop Nor (China), Maralinga (Australia) and others in the Pacific islands, India, Pakistan and Korea. The first atmospheric tests were conducted at the Nevada test site (USA) in 1951. Following the Limited Test Ban Treaty of 1963, atmospheric testing ceased, and nearly 90 percent of the US underground weapons tests were detonated in Nevada. Congress imposed a moratorium on testing of nuclear weapons, and in 1992, under­ground testing ceased. A total of 907 underground nuclear detonations were conducted above, near and below the groundwater table in alluvial basins, in volcanic highlands, in shafts and tunnels of zeolitised volcanic rocks, and in tunnels mined in granitic rock. Underground testing at Nevada deposited an estimated 132 million curies of radioactivity below ground, decay cor­rected to 1992. These topics are considered in details in the last three chap­ters of this book.

An underground explosion produces a spherical cavity from combined vaporisation, melting and shock compression of the rock. As the detonation pressure subsides, the rocks above the cavity typically collapse (timeframe of seconds to days after the test) and the cavity is filled with rubble consist­ing of collapsed rock, and solidified rock melt (melt glass). The collapse void can propagate upward variable distances forming a chimney that may or may not extend to the surface forming a subsidence crater. The temperature and pressure history of an explosion and response of the surrounding rock control the distribution of radionuclides around the test. Radionuclides produced underground include tritium, fission products, actinides and acti­vation products. Refractory radionuclides are trapped primarily in the melt glass, and in cavity rubble and compressed rock around the cavity; volatile species circulate outward and condense in cracks and void spaces for dis­tances of 1-3 cavity radii from the test point (Pawloski et al., 2008). The extensive contamination of the land at such sites and the potential for spread via local hydrology and hydro-geological has led to extensive studies of such sites (e. g., Busygin et al., 1996; D’Agnese et al., 1997).