This section primarily discusses non-thermal methods of encapsulation. The thermal encapsulation by glass is covered in Section 6.4.1 on glass ceramic materials on pages 214-217.

Cements including grouts

Stabilization and solidification with cement-based binders has been used to immobilize radioactive wastes since the beginning of the nuclear age. The process has been used to encapsulate solid waste, solidify liquid waste (including tritiated water), stabilize contaminated soils, stabilize tank-heel residues after tanks are emptied, and as low permeability barriers. Cements have also been used as binders and to encapsulate granular or cracked waste forms.

Cements microencapsulate wastes, although there is recent evidence that during hydration three binding mechanisms can also occur between the cement and metal ions in the waste [99-101]:

• precipitation of metal ions into the alkaline matrix as an oxide, mixed oxide, or as another discrete solid phase;

• adsorption or (co-)precipitation of metal ions onto the surface of cement minerals;

• incorporation of metal ions into hydrated cement minerals as they crystallize.

These mechanisms are shown as examples in Table 6.9; with the binding mechanisms (reaction of the waste with the cement or grout particles) shown as encapsulation and without the binding mechanism shown as embedding.

These processes are not mutually exclusive (so both encapsulation and embedding take place) and the above classification partly reflects slow kinetics; previously adsorbed species may be incorporated as mature cement pastes. Nevertheless, it does allow some generalized guidelines to be for­mulated. The solubility of discrete heavy metal solid phases is a limiting factor with regard to the second and third mechanisms [102], so that only ions that do not precipitate as basic oxides tend to be incorporated in, or surface adsorbed to, hydrated cement minerals to a significant degree.

The principal minerals available in the hydrated Portland cement matrix are calcium silicate hydrate (C-S-H, 50 wt%), portlandite (Ca(OH)2, 20 wt%), and Ca aluminates. The most important Ca aluminates
are ettringite (3CaO. Al2O3.3CaSO4.32H2O, 4 wt%), calcium aluminate monosulphate (3CaO. Al2O3.CaSO4.12H2O, 7 wt%) and Ca carboaluminate (3CaO. Al2O3.CaCO3.11H2O, 7 wt%) [103] . Together they make up almost 90 wt% of the mineral suite in hydrated ordinary Portland cement (OPC) paste and thus, have the greatest potential for metal(loid)-ion binding. The relative importance of the above processes for selected metals can be found in a recent review [104].

OPC is the most common type of cement used for immobilizing liquid and wet solid wastes worldwide [6]. Composite cement systems were devel­oped in the UK for ILW encapsulation using additional powders as well as OPC such as blast furnace slag (BFS) and pulverized fuel ash (PFA). These offered cost reduction, energy saving and potentially superior long-term performance. BNFL, for example, use a 9:1 ratio of BFS to OPC to reduce the heat of hydration, which for OPC cements, would otherwise limit con­tainer volumes. Large containers (see Fig. 6.3) can therefore be used safely without concern over heat from setting reactions causing water to boil off.

Modeling has shown that cements can be ‘designed’ to retain radioactive and hazardous constituents [ 105] . In fact, much research has focused on improving the effectiveness of grout in adverse environments associated with the disposal of radioactive waste [106-108]. As discussed in these refer­ences, a variety of cement-polymer composites have been investigated as a means of making grouts more compatible with the radioactive and chemical constituents in waste.

For example, the addition of blast furnace slag to the Saltstone cement[20] being used to solidify Cs-decontaminated salt supernate at the Savannah River Site (SRS), provides a chemical reductant [iron(II)] and a precipitat­ing agent (sulfide) that chemically binds contaminants such as chromium and technetium as insoluble species, thus reducing their tendency to leach from the waste form. Experimentation has shown that leaching of chro­mium and technetium was effectively reduced to levels that would allow all projected future salt solution compositions to be processed into Saltstone [109] . Long-term lysimeter studies have shown that the addition of slag essentially stopped technetium-99 leaching, although it did not reduce nitrate leaching [109]. Because the SRS Saltstone admixture that is blended with 45% liquid waste is only 10 wt% OPC, 25 wt% fly ash, and 25 wt% slag, it is a geopolymeric cement as the alkali in the salt supernate reacts with the fly ash in geopolymer-like chemical reactions.

The water in the hydrated cement blends may generate H2 by radiolysis in high radiation fields and require vented canisters [110] when container­ized. While this study concentrated on transuranic (TRU) wastes containing 238Pu oxide, which is primarily alpha radiating, the other studies have dem­onstrated the radiolysis of concrete with “Co (gamma radiation) and [H (beta radiating) [111-113].

Recent comprehensive reviews of cement systems for radioactive waste disposal can be found in Pabalan et al. [114] and Glasser [115]. Long-term cement durability comparisons have been made using ancient cements, geopolymers, and mortars [116-123], some of which may also serve as natural analogues for geopolymer wasteforms [124, 125].

The cements and grout formulations are too extensive to list as examples. The durability response is complex due to the relative response of encap­sulation with some chemical reaction and embedding. Therefore, the dura­bility is usually modeled as a diffusion rate with respect to the element(s) of interest.

The various stages of the nuclear fuel cycle and the operation and decom­missioning of nuclear reactors all have the potential to create contaminated sites. The contamination may include spillage of ore end-product at the mine and in transport; waste from enrichment and fuel fabrication opera­tions; fission product and actinide waste streams from reprocessing of fuel elements; radioactive effluents from normal operations of nuclear power plants; wastes produced during decommissioning of reactors; and major releases under accident conditions.

Russia adheres to the principles of safe RAW management [18] , involving the following basic stages of the process:

1. Pretreatment

• collection

• segregation

• chemical adjustment

• decontamination.

2. Treatment

• volume reduction

• removal of radionuclides from the waste

• change of composition.

3. Conditioning

• solidification

• immobilization

• overpacking.

RAW is divided into categories based on its origin; its physical, chemical and biological properties; its state of aggregation; and level of activity. The different properties of different types of RAW make it necessary to use different RAW management technologies for different sorts of waste. In Russia, a number of technological methods have been developed to allow the optimal processing of RAW. The basic methods used for RAW process­ing [19] are presented in Table 10.5 and Fig. 10.5.

LRAW constitutes the majority of all RAW in Russia, with 90% of LRAW made up of aqueous solutions originating from (a) technological drains, which are produced at industrial and research centres, including medical and biological laboratories; and (b) decontamination drains from the decontamination of equipment and overalls. These aqueous solutions are highly varied in their chemical and radionuclide composition, and include such diverse forms as ions, dissolved complexes of organic sub­stances, colloidal particles and micelles, suspended solids, and liquid emulsi­fied oil products. Solid RAW includes different materials contaminated with radionuclides such as construction materials, dismantled equipment, spent filters and resin, corpses of experimental animals, and silts and soils from contaminated territories.

The basic aim of RAW treatment is to reduce the physical volume of the waste and to transfer this waste into a monolithic, chemically and mechani­cally stable form, suitable for long-term storage in containers.

10.5 Basic methods of RAW treatment at MosNPO ‘Radon’ [19].

A. V O K A L, Radioactive Waste Repository Authority, Czech Republic and P. STOCH, Institute of Atomic Energy, Poland

DOI: 10.1533/9780857097446.2.415

Abstract: The chapter describes radioactive waste (RAW) issues in the Czech Republic, Slovak Republic and Poland. The situation in the Czech and Slovak Republics is different from Poland. Poland has run only experimental reactors, while the Czech and Slovak Republic have operated nuclear power plants (NPPs) since the 1970s. The Czech and Slovak nuclear programmes were based on the assumption of returning reactor spent fuel (SF) assemblies to the Soviet Union without any commitment concerning SF destiny. After the decision of the Russian Federation to cancel ‘free of charge’ returning to the Russian Federation, both countries started to develop their own concept concerning SF di. sposal. The main problem for Poland is that their repository at Rozan for RAW from industry, medicine and research is almost full and it necessitates finding a new facility for accepting waste by 2020.

Key words: radioactive waste (RAW), Czech Republic, Slovakia, Poland, spent fuel disposal.

12.1 Introduction

This chapter is devoted to the description of the radioactive waste (RAW) management situation in the Czech Republic, Slovak Republic and Poland. The situation in the Czech and Slovak Republics is different from Poland: Poland has not yet run any nuclear power plant (NPP), and is only consider­ing starting its first NPP by 2020, whereas both the Czech and Slovak Republics have operated NPPs since the 1970s. Poland does, however, operate a number of experimental nuclear reactors. Figure 12.1 shows the locations of nuclear installation in the Czech Republic.

All of these countries are considering building new NPPs. Today in the region, public acceptance of nuclear energy is quite high (generally about 60%, in some countries up to 70%). Nevertheless, because of the possible consequences far beyond national borders in case of an accident, and because nuclear energy is also such a divisive issue among various

12.1 Locations of nuclear installations in the Czech Republic.

opponents, there is no doubt that, for each project to build new NPPs will require full and frank information regarding three key issues to be set out (IAEA, 2009):

• nuclear safety,

• non-proliferation, and

• RAW and spent fuel (SF) management.

Lessons learned from building and operating RAW and SF management systems in these countries, as described in this chapter, may, therefore, sig­nificantly contribute to preparing new, improved systems of RAW and SF management already included in the designs of the new NPPs.

In the case of the Czech Republic, this chapter will focus mainly on sum­marising information from RAW management at two NPP with Russian WWER 400 and WWER 1000 reactors located at Dukovany and Temelin. Less attention will be devoted to RAW management from the use of ion­ising radiation in industry, medicine or research including SF management from research reactors. This is because, compared with waste from NPPs, waste from other sources is not so great a problem in the Czech Republic and management of this sort of waste was established in the 1960s and suit­able disposal facilities are available. However, problems with remediation
of contaminated sites after extensive uranium mining and milling in the Czech Republic will be highlighted.

Decommissioning of nuclear facilities, in addition to RAW management at Jaslovske Bohunice and Mochovce NPPs with WWER 440 reactors, is a big issue in Slovakia, because of an operational incident at the first Czecho­slovak NPP (A1) in 1977, after which it was shut down. In addition, the closure of the first generation of WWER reactors (V-230 type) at NPP (V1) at Jaslovske Bohunice was one of the conditions for fulfilling the Accession Agreement of Slovakia to the European Union.

The major topic for Poland concerning RAW management is primarily disposal of SF assemblies from research reactors at Swierk and manage­ment of radioactive waste from industry, medicine or research.

For practical purposes, radioactive waste is classified into different classes depending on actual management needs. A number of parameters are con­sidered in classification schemes, the most important of which are shown in Table 1.1 .

Radioactive wastes are typically classified accounting for potential clear­ance, decay storage or disposal, e. g. final point of waste disposition (IAEA, 2009). Key parameters in the IAEA (International Atomic Energy Agency) classification scheme are radionuclide half-life and radioactivity content. The radionuclides are divided into long-lived and short-lived, where a radi­onuclide with a half-life longer than that of 137Cs (30.17 years) is considered to be long-lived, whereas those with shorter half-lives are considered short­lived. The activity content is a generic term that covers activity concentra­tion and total activity and is used in classification schemes accounting for the generally heterogeneous nature of radioactive waste (IAEA, 2009). The activity content can range from negligible to very high, e. g. very high con­centration of radionuclides or very high specific activity. The radioactivity contents are always analysed compared to exemption levels (IAEA, 2004), e. g. the higher the activity content above those levels the greater the need to contain the waste and to isolate it from the biosphere.

The IAEA classification is shown schematically in Fig. 1.4 and has as lowest by activity content the exempt waste (EW). Exempt waste (EW) is

Property Origin Radiological Physical Chemical Biological Parameter Source, Criticality. Physical state (solid, Potential chemical Potential hazard. Manufacturer Half-life. Heat generation. Intensity of radiation. Activity and concentration of radionuclides. Surface contamination. Dose factors of relevant radionuclides. liquid, gas). Size, volume and weight. Compressibility Dispersibility. Volatility. Solubility. Miscibility. hazard. Corrosion resistance. Corrosivity. Organic content. Combustibility. Reactivity. Gas generation. Sorption of radionuclides. Decomposition rate and products.

that radioactive waste that meets the criteria for clearance, exemption or exclusion from regulatory control for radiation protection purposes which are given in IAEA publications (IAEA, 2003b, 2009). The criteria for exemption were established by the IAEA following the ICRP (Interna­tional Commission on Radiological Protection) recommendations and prin­ciples used to derive exemption levels for radioactive materials. Generically they are based on an expected individual effective dose not higher than 10 pSv/annum and a collective effective dose not higher than 1 person Sv/ annum. Exemption levels were established for both concentration and total amount of radionuclides based on the individual and collective dose. These were determined for each radionuclide taking account of all possible path­ways to humans including assessment of individual and collective doses. Exemption levels are published in the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (IAEA, 2003b). Sources of radiation are exempt from control if at a distance of 0.1 metres, the dose rate is below 1 pSv/h. Clearance levels are defined by the national regulatory authorities; however, since these take into account internationally approved recommendations, quantified clear­ance levels (with some exceptions) are similar in all countries. EW contains such small concentrations of radionuclides that it does not require provisions for radiation protection, irrespective of whether the waste is disposed of in conventional landfill sites or recycled, so EW is, in practice, considered as a non-radioactive material.

The IAEA classification scheme defines five classes of radioactive waste: 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 EW, but that does not need a high level of containment and isola­tion and because of that 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. A precise boundary between LLW and ILW cannot be universally provided, as limits on the acceptable level of activity concentration will differ between individual radionuclides or groups of radionuclides. 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 4,000 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 also be 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 recognised HLW disposal option.

As can be seen, the IAEA classification scheme is rather generic and has no exact limits in defining radioactive waste classes. Existing national regu­lations give more exact figures (Ojovan and Lee, 2005). In the UK, radioac­tive wastes are classified as VLLW, LLW, ILW and HLW (Table 1.2).

High level waste is waste with levels of activity and radionuclide concentra­tions high enough to generate significant quantities of heat by radioactive decay or waste with large amounts of long-lived radionuclides that need to be considered in the selection of a disposal facility and disposal route for such waste. Handling and storage of HLW requires proper shielding and in some cases also additional cooling. Typical examples of HLW generated at NPPs are highly activated reactor parts. However, the main source of HLW is reprocessing of spent nuclear fuel. Disposal in deep, stable geological formations usually several hundred metres or more below the surface is the generally recognized option for disposal of HLW.

An internationally endorsed, non-binding Code of Conduct [50] was approved in 2004 to facilitate the safe management of radioactive sources, including disused sealed sources that in most cases are declared as RAW. The objectives set out in the code should be achieved through the establish­ment of a comprehensive system of regulatory control of sources, applied from their initial production to their final disposal, and a system for the restoration of such control if it has been lost. To facilitate the implementa­tion of the Code of Conduct, in 2005 a document ‘Guidance on the Import and Export of Radioactive Sources of Category 1[7] and Category 2[8]’ was also agreed 651]. As of May 2011, 103 Member States have expressed support for the provisions of the Code of Conduct [52].

Stainless steel hardware (including spacers, clips, springs, end plates, etc.) is removed in the first stage of reprocessing. The primary radionuclides in the hardware come from activation of stainless steel components (e. g., 54Mn, 55Fe, 60Co, and 63Ni). This stream makes up roughly 5 wt% of the fuel assem­bly. Cladding (sometimes referred to as hulls) refers to the metal tubes used to hold the fuel. At present, cladding compositions are typically >95% Zr with Sn and/or Nb alloying agents although stainless steels are sometimes used (e. g., in UK advanced gas-cooled reactor fuel). The primary radionu­clides in and on the cladding are activation products (e. g., 95Zr, 95Nb, 54Mn, 55Fe, 60Co, and 125Sb), tritium, and transuranics/fission products from alpha recoil at the fuel-cladding interface. Tests have shown that the transuranic contamination resides in the inner 7 pm of the cladding. As cladding makes up roughly 25% of the used nuclear fuel mass, it is the largest single fuel component after UO2.

Traditionally, hulls and hardware have been managed together. The most common approaches to managing these wastes are to wash and then (1) embed them in cement for disposal, (2) dispose directly, and (3) compact and dispose (IAEA, 1985). Compared to direct disposal, the compaction reduces disposal package volume by roughly a factor of four, while encap­sulation in cement increases the volume by roughly 100% (double the volume).

A number of alternative approaches have been studied for these wastes including: rolling compaction and cementation, embedding in graphite, compaction with malleable metals (e. g., Pb), compaction and encapsulation in low temperature metals, powder metallurgical encapsulation, glass encap­sulation, cryogenic crushing and encapsulation, oxidation and conversion to ceramic waste forms (e. g., zircon), oxidation and cementation, hot press­ing, melting to a Zr-Fe alloy zirconium separation using reactive gasses (Collins et al., 2011). Although these methods are not being currently imple­mented, many show promise for improved waste management compared to the reference technologies.

T. M. AHN, US Nuclear Regulatory Commission, USA

DOI: 10.1533/9780857097446.1.273

Abstract: Examples of analytical approaches and methodologies for modelling the behaviour of waste forms and waste package metals in long-term management of spent nuclear fuel (SNF) and high level waste (HLW) are presented. Two cases, long-term geological disposal and interim extended dry storage, are considered. The integrity of the waste package (or canister) that serves as a barrier is dependent upon the performance of construction metals. Corrosion degradation modes of the construction metals are evaluated. The waste behaviour during SNF degradation is also evaluated. In each mode of corrosion or degradation, the associated risk insights are discussed in the system performance of disposal or storage.

Key words: assessment and modelling, nuclear waste form, nuclear waste package, storage, disposal.

7.1 Introduction

This chapter presents example analytic approaches and methodologies for modelling the behaviour of waste forms and different metals used in pack­aging spent nuclear fuel (SNF) and high level waste (HLW). The long-term behaviour of waste forms and different metals used for packaging SNF and HLW are important attributes in assessing safety and security associated with nuclear waste management. This behaviour is a core component in determining radionuclide source-term and/or criticality control, used in assessing radionuclide release to the human environment. The assessments and modelling of the long-term behaviour of the waste form and different metals are further complicated by a variety of environmental conditions, including natural and human-induced external hazards. This is especially true when the purported waste management time is very long, e. g., several thousand years or beyond.

Disclaimer: The NRC staff views expressed herein are preliminary and do not constitute a final judgment or determination of the matters addressed or of the acceptability of any licens­ing action that may be under consideration at the NRC.

The approaches and methodologies presented in this chapter also cover model uncertainties that affect assessment of public health and safety. The content of this work is considered by the US Nuclear Regulatory Commis­sion (NRC) which assesses cases for management of SNF and HLW in the US. The NRC prepares risk and performance insights. Information that the NRC obtained from the past activities in the management of SNF and HLW in the US, along with relevant information from different international programs, is included.

In the US, the long-term management of SNF and HLW is considered for geological disposal for thousands of years and beyond, and interim extended dry storage of SNF for up to 300 years. In both management cases, materials performance issues related to waste form and corrosion of differ­ent construction metals are considered, i. e., container metal in waste pack­ages used for geological disposal, and canister construction metal for extended dry storage. The following four topics related to the waste-form dissolution and corrosion are addressed in this chapter: long-term integrity of passive film, slow general corrosion, and localized corrosion of different metals; stress corrosion cracking (SCC) of carbon steel and stainless steel; SNF degradation; and cladding performance. The discussion of each topic addresses how it applies to the two management cases, as appropriate. Broader performance issues of waste form and different metals under the two management cases are also discussed. Finally, risk insights are addressed with respect to performance (or risk) assessment for the disposal or storage system. Both management cases incorporate laboratory data, analytical models, archaeological (for disposal) and/or industrial (for extended dry storage and disposal). Some similar classes of metals are used in both cases for different purposes. For example, stainless steels are primarily used for extended storage, but may also be considered for disposal. On the other hand, carbon steel is mainly applied in disposal.

In-situ remediation technologies for control or treatment of soils and groundwater are increasingly being investigated because they offer the potential for:

• significant cost reduction of clean-up by eliminating or minimizing exca­vation, transportation, and disposal of waste;

• reduction of health impacts on workers and the public by minimizing exposure to waste during excavation and processing;

• significant reduction in ecological impacts, and

• remediation of inaccessible sites, including deep sub-surfaces and in, under, and around buildings.

In-situ technologies can be subdivided into five major groups:

1. Containment technologies (e. g., bottom sealing, surface capping, polymer concrete barriers, cryogenic barriers, fluidized-bed zeolite system, plasma arc glass cup, slurry wall, soil/cement wall, vitrified barriers).

2. Solidification and stabilization (e. g., lime-fly ash Pozzolan system, organic binding, Pozzolan-Portland cement system, sorption, in-situ encapsulation, in-situ compaction).

3. Physical-chemical treatment (e. g., de-chlorination, electro-acoustics, electro-kinetics, neutralization, oxidation/reduction, precipitation/floc — culation, soil flushing, in-situ steam/air stripping, simultaneous injection, extraction and recharge, vacuum extraction).

4. Thermal treatment (e. g., radio frequency and electromagnetic heating, in-situ vitrification).

5. Biological treatment (e. g., biomass remediation, biodegradation).