H. RIN D O, K. TAKAHASHI and M. TACHIB ANA, Nuclear Cycle Backend Directorate, Japan

DOI: 10.1533/9780857097446.2.723

Abstract: This chapter summarizes the current strategy and policy for radioactive waste management in Japan which has been hindered by a lack of public acceptance and of a final high level waste end-point (geological repository). Ongoing decommissioning of several nuclear facilities, including the Tokai-1 NPP, the Advanced Thermal Reactor ‘Fugen’ and the Plutonium Fuel Fabrication Facility (PFFF), are described.

Key words: radioactive waste treatment, radioactive waste disposal, decommissioning and dismantling, nuclear facilities, policy and strategy, Japan.

22.2 Introduction

This chapter was written largely before the Fukushima accident, details of which and the clean-up programme are included in the next chapter. However, not only nuclear policy but also nuclear safety regulation in Japan is likely to change in the future.

On December 16, 2011, TEPCO confirmed that the release of radioactive materials was under control and that radiation doses were being signifi­cantly reduced [5]. In April 2012, the predicted equivalent radiation doses per year for areas near the Fukushima Daiichi NPP for the next 20 years were released [20]. As shown in Plate XIII (between pages 448 and 449), a dose of more than 100 mSv/yr may still be encountered about 23 km north­west of the plant until March 2013, and the 50 mS/yr dose zone can only be confined to a 20 km radius after March 2017. Since the standard worker dose limit for Japanese workers is 50 mSv per year and 100 mSv over 5 years [21] , certain areas will still be subject to high alert for radiation effects in the near future. By March 2022 (11 years after the accident), certain hot spots may still possess a radiation dose higher than 50 mS/yr. These hot spots will mostly be eliminated after 20 years, as shown in the prediction for March 2032 in Plate XIII.

V. P. B U S Y GIN, Defence Department, Russia

DOI: 10.1533/9780857097446.3.833

Abstract: This chapter reviews and discusses the effects of residual features on the long-term geothermal activity in the epicentral zone of underground nuclear explosions (UGE). The thermal anomaly parameters and their connection to carrying out thermal surveys and surface thermal logging on the present day surface are determined. A remote method of measuring the thermal anomalies is proposed.

Key words: underground nuclear explosion, radioactive waste, thermal radiation, monitoring, epicentral zone.

27.1 Introduction

Worldwide, 2,054 nuclear explosions have been conducted since 1945, including 1,524 underground explosions (many explosions were carried out in groups) (Kochran et al., 1992; Mikhailov, 1992, 2001). The last 1,373 explo­sions were performed at special nuclear test sites:

• 333 explosions at Semipalatinsk and West Kazakhstan (former Soviet territory, at present the territory of the Republic of Kazakhstan),

• 39 explosions at Novaya Zemlya (Russia),

• 781 explosions in the US (Nevada),

• 3 explosions in the USA on the Island of Amchitka (Alaska) landfills,

• 13 explosions in Algeria (District Hoggar),

• 147 explosions on the islands of Mururoa and Fangataufa (France)

• 24 explosions at the Nevada test site in the US were performed by the United Kingdom,

• 24 explosions were performed at the Lop Nor test site in China,

• Miscellaneous test explosions were carried out by India (3), Pakistan (2), and North Korea (2) (Mikhailov, 2001).

Other underground nuclear explosions were carried out underground at various test sites or on the surface, but with the purpose of applying the technology of nuclear explosions for peaceful solutions of a variety of tech­nical problems (Mikhailov, 2001; Logachev et al., 2001).

In the Soviet Union from 1961 to 1987, in accordance with Programme No 7 ‘Nuclear explosions for the national economy,’ 124 industrial com­plexes experienced an explosion, of which a number were carried out at the Semipalatinsk test site. Outside the territory of the present-day Russia, 80 explosions were carried out in the Republic of Kazakhstan (outside the polygon), 32 in the Ukraine, two in Uzbekistan, and two in Turkmenistan. The majority of the explosions were carried out in camouflet option, i. e. without a breakthrough cavern explosion into the atmosphere, and were aimed at solving problems: seismic sensing (39), creation of industrial con­tainers for food storage (26), working out the technology and scientific experiments (22), intensification of oil fields (21), eliminating emergency fountain (5), creating reservoirs (4), waste disposal in deep horizons (2), crushing ore (2), prevention of gas emission in coal seams (1), creating channels (1), and tailings dams (1) (Mikhailov, 2001; Israel, 1974).

Most of the explosions were carried out under difficult physical and geological conditions: in permafrost, semi-deserts, mountains, and salt for­mations in mining areas. Together with the explosion parameters and the monitoring information, these conditions determine the nature of residual geophysical phenomena, i. e. cleavage zones, zones of increased fracturing, changes in the permeability induced by electric and magnetic fields, thermal effects, and possible contamination with radioisotopes, which are precursors of volatile radioactive elements, increased release of radioactive radon gas, and changes in the environmental performance of the natural environment, etc.

This chapter describes the features and control areas of underground nuclear explosions and potential changes over long time periods, which allow evaluation of the state of the environment, i. e. the outward manifesta­tion of certain physical fields on the surface.

Section 27.2 describes the basic mechanisms of the boiler cavity, pillar collapses, and the cleavage phenomena on the surface, while also summariz­ing the classification and spatial distribution of radioactive waste. Section 27.3 examines the long-term problematic situations that arise at the surface, in aquifers and hydrocarbon horizons in the zone of underground nuclear explosions. There are cases that require regular monitoring. Section 27.4 is devoted to describing the results of thermal imagery and ground tempera­ture well logging in areas of underground nuclear explosions. A phenom­enological model of formation and dynamics of thermal anomalies is developed. Links are made between thermal anomalies, the level of gamma background radiation, and radon releases. In Section 27.5 we propose a method using monitoring by spacecraft to measure thermal anomalies. The prospects of applying this method for global monitoring of the effects of underground nuclear explosions are determined.

The commercial electric power reactors in Finland and Sweden generate by far the majority of the radioactive waste (RAW) in the Nordic countries. The waste is classified into three categories: operational waste or reactor waste, decommissioning waste and spent nuclear fuel.

The treatment of the waste depends on the activity level. The operational waste, which accounts for about 85% of all wastes from the reactor opera­tions, consists mostly of low and intermediate level waste that requires isolation from the environment for at least 500 years. The low level waste (LLW) can generally be handled without radiation shielding, while some shielding will be necessary for the intermediate level waste (ILW).

The decommissioning waste is mostly the scrap metal and concrete from the future dismantling of the reactors. Most of the waste will be LLW and ILW, but the reactor pressure vessel and its internal components are clas­sified as long-lived waste and must be isolated for thousands of years.

The spent nuclear fuel is only a small fraction of the waste, but it is the most dangerous waste that must be isolated for very long time periods.

In addition to these major waste sources, there is also waste from research and industrial and medical applications. This will also include some research reactor fuel. For some research reactor, e. g., Studsvik and Ris0, however, the fuels were on loan from the US and have been shipped back there.

Radioactive waste has been generated in increasingly large amounts since the advent of the nuclear industry in the 1940s and 1950s. Contaminated environments have also been experienced from that era; a considerable number since the start of uranium mining, some from military-related activ­ities — both from processing plants associated with weapons production and weapons testing and from nuclear accidents, in particular Chernobyl and more recently Fukushima. The amount of RAW generated to date in non­military programmes is generally reported in the open literature, but that from military activities can only be estimated from weapons production activities. A review of the global inventory has been made by the IAEA and is summarised in Table 3.1 [4].

aln reality a relatively minor fraction of the spent fuel generated by nuclear power plants (NPPs) has been reprocessed and has been transformed into a variety of products, including different classes of radioactive waste.

b A fraction of the HLW generated by reprocessing civilian spent fuel has been vitrified. Most HLW generated by defence programmes is stored in liquid form.

“Estimates are highly uncertain. In some countries there is no clear separation between reprocessing for military and for civilian purposes.

Conditioning includes those operations that produce a waste package suit­able for handling, transport, storage and/or disposal [11]. Conditioning may include the conversion of the waste to a solid waste form, additional immo­bilization of some solid waste, packaging of the waste form into containers, and, if necessary, providing an overpack. The waste form is the waste in its physical and chemical form after treatment and/or immobilization (result­ing in a solid product) prior to packaging. The waste form is a component of the waste package. The immobilization of RAW (solidification, embed­ding or encapsulation) to obtain a stable waste form is an important step in waste management needed to minimize the potential for migration or dispersion of radionuclides into the environment during storage, handling, transport and disposal. Radioactive and chemically hazardous constituents in the waste can be immobilized into a waste form material through two processes: Constituents can be (1) bound into the material at atomic scale (chemical incorporation), or (2) physically surrounded and isolated by the material (encapsulation) [11].

A number of matrices have been used for waste immobilization and those include glass, ceramic, cement, polymer and bitumen [11, 12, 21, 29-36]. The choice of the immobilization matrix depends on the physical and chemical nature of the waste and the acceptance criteria for the storage and disposal facilities to which the waste will be consigned. Several factors must be con­sidered when selecting a waste form material for immobilizing a specific waste stream. The key considerations include the following [11, 21, 33, 34]:

• Waste loading: The waste form must be able to accommodate a signifi­cant amount of waste (typically 25-45 w%) to minimize volume, thereby minimizing the space needed for storage, transportation and disposal.

• Ease of production: Fabrication of the waste form should be accom­plished under reasonable conditions, including low temperatures and, ideally, in an air atmosphere, using well-established methods to mini­mize worker dose and the capital cost of plant.

• Durability: The waste form should be physically durable and have a low rate of dissolution when in contact with water to minimize the release of radioactive and chemical constituents.

• Radiation stability: The waste form should have a high tolerance to radiation effects from the decay of radioactive constituents. Depending on the types of constituents being immobilized, the waste form could be subjected to a range of radiation effects, including ballistic effects from alpha decay and ionizing effects from decay of fission product elements.

• Chemical flexibility : The waste form should be able to accommodate a mixture of radioactive and chemical constituents with minimum forma­tion of secondary phases that can compromise its durability.

• Availability of natural analogues: Since direct laboratory testing of the waste forms over the relevant timescales for disposal (typically lOMO6 years) is not possible, the availability of natural mineral or glass ana­logues may provide important clues about the long-term performance of the material in the natural environment, thereby building confidence in the extrapolated behaviour of the waste form after disposal.

• Compatibility with the intended disposal environment: The waste form should be compatible with the near-field environment of the disposal facility. The near-field environment provides the physical and chemical conditions that are favourable for maintaining waste form integrity over extended periods, which helps to slow the release of constituents and their transport out of the facility.

A number of materials have been used for waste immobilization and those include glass, ceramic, metal, cement, polymer and bitumen. All these ma­terials have their advantages and disadvantages both in terms of the kinds of waste that can be immobilized and the properties of the solidified waste forms obtained. The choice of the immobilization matrix depends on the physical and chemical nature of the waste and the acceptance criteria for the disposal facility to which the waste will be consigned.

Glass is being used worldwide to immobilize HLW from reprocessing of spent nuclear fuel and targets [11, 21, 35] : The immobilization process, vit­rification, is a continuous process capable of handling large-volume waste streams. Vitrification has demonstrated its efficiency and its flexibility in a number of countries. It has become the reference process for the condition­ing of HLW and is currently deployed on a large scale for lower activity waste streams. Given the good results obtained with vitrification of HLW, several projects are underway for the vitrification of slurries, low and inter­mediate level solid waste, mixed waste, etc. The advantages are: volume reduction, destruction of organic constituents including hazardous materi­als, immobilization of radioactive and hazardous components, advantages for storage, transportation and disposal. The vitrification processes are suf­ficiently robust, which means that they accept almost any waste after a minimum of up-front characterization, with reproducible characteristics of the end product and acceptable off-gasses. Vitrification can also be per­formed in situ as a special case (e. g., legacy waste or contaminated soil) [23].

Crystalline ceramics are inorganic, non-metallic solids that contain one or more crystalline phases. Single-phase crystalline ceramics can be used to immobilize separated radionuclides (e. g., plutonium-239) or more chemi­cally complex waste streams (e. g., HLW) [11, 36] . In the latter case, the atomic structure of the ceramic phase must have multiple cation and anion sites that can accommodate the variety of radionuclides present in the waste stream. These materials are potentially attractive for immobilizing long — lived alpha-emitting actinides such as plutonium, neptunium and ameri­cium. However, some of these materials are susceptible to radiation damage effects associated with alpha decay from actinides [36]. Multiphase crystal­line ceramics (e. g., Synroc) consist of an assemblage of crystalline phases. Individual phases are selected for the incorporation of specific radionu­clides, with the proportions of phases varying depending on the composition of the waste stream. An individual phase can host one or more radionu­clides, including solid solutions of radionuclides. However, not all phases will host radionuclides. Ceramic materials and methods of fixation are largely at an early stage of developement. Ceramic products are crystalline in nature and therefore thermodynamically stable although they are sensi­tive to radiation damage.

Glass-composite materials (GCMs) are materials that contain both crys­talline and glass phases [11, 21, 34, 35]. Depending on the intended applica­tion, the major component may be a crystalline phase with a vitreous phase acting as a bonding agent. Alternatively, the vitreous phase may be the major component with particles of a crystalline phase dispersed in the vitre­ous matrix. GCMs can be formed by a number of processes, including melt crystallization (controlled or uncontrolled), multiple heat treatments, or by encapsulation of ceramic material in glass. GCMs offer several potential advantages over glass for use as waste form materials, including increased waste loadings, increased waste form density, and thus smaller disposal volumes. These waste forms can also be used to immobilize glass-immiscible components such as sulphates, chlorides, molybdates, and refractory materi­als that have very high melting temperatures. They can also be used to immobilize long-lived radionuclides (e. g., actinides) by incorporating them into the more durable crystalline phases; short-lived radionuclides (e. g., many fission products) can be accommodated in the less durable vitreous phase [35] .

Cements are inorganic materials that set and harden as a result of hydra­tion reactions [11, 21]. Cements are used to immobilize waste having relatively low levels of radioactivity (i. e., low or intermediate level RAW). Higher activity wastes can result in radiolysis and production of hydrogen gas from the breakdown of water or hydroxyl groups in the cement. Cementation is viewed as low cost and forms a major part of both solid and liquid (mainly aqueous) LLW and ILW immobilization technolo­gies. The range of applicability of cements is to be considered in view of the characteristics of the environment and of the initial waste. The cement may display pH buffering properties and, consequently, control mobility of most radionuclides in the disposal environment. The quality of cemented waste forms is continually being improved. They are efficient for the immobiliza­tion of alpha-bearing waste. The problematic side is the potential for reac­tion with some wastes and the relatively high porosity and leachability for some radionuclides of the end-product. In addition, cementation results in a volume increase rather than decrease, so a smaller than unit reduction factor.

Bitumen, a viscous hydrocarbon and a major component of asphalt, has been used to solidify and stabilize radioactive materials [11, 32] , Bitumen immobilizes waste mainly by encapsulation rather than binding the waste chemically. The advantages of bitumen as a waste form are simplicity of production, low operating cost, and leach-resistant characteristics. However, bitumen can be a fire hazard, especially when oxidizing wastes like nitrates are involved.

Some waste treatment methods, such as plasma arc melting, or molten metal techniques, result in both a high volume reduction and very stable waste forms.

When spent fuel is not reprocessed (the once-through fuel cycle), SFW conditioning consists of volume optimization (rearrangement of the fuel rods) and enveloping them in a multi-component barrier consisting of various metals (copper, lead) and the packaging canister [25] , Several dif­ferent types of metallic materials have been studied as potential waste forms. Like crystalline ceramics, metal waste forms can consist of single — or multiple-phase assemblages and the waste form itself can be granular or monolithic. Metal waste forms can be fabricated by sintering or casting. Each of these techniques has drawbacks; in particular, it can be difficult to find metal compositions and processes that effectively wet and encapsulate dispersed phases or fines.

The major types of waste forms will be described in regard to the manner in which the radionuclides are immobilized and the methods by which each can be made. Different waste forms give different durability tests responses. Single-phase waste forms (glass and single-phase oxides or crystalline ceramics (minerals) have only one source of radionuclides that can leach during a durability test. In multiphase waste forms the distribution of the radionuclides amongst the phases present becomes important as each phase has its own rate of leaching for the specific elements that it sequesters. Each waste form given in Tables 6.3-6.10 will be described in terms of the radio­nuclide immobilization achieved and references given as to which condi­tioning technologies can be used to make each type of wasteform.

The immobilization of HLW is always achieved by its atomic-scale incor­poration into the structure of a suitable matrix (typically glass, a GCM, or a crystalline ceramic (also sometimes referred to as mineral analog waste forms) so that the radionuclides are incorporated into durable structures by any combination of short range order (SRO),[13] medium range order (MRO)[14] [15] or long range order (LRO).2 Glasses incorporate radionuclides and hazardous species into their atomic structure by SRO and MRO [16]. Recent experimentation has shown the existence of large cation-rich clus­ters in glass, e. g. clusters of Ca in CaSiO3 glasses and clusters of Na2MoO4

Waste form Homogeneous glass	Inhomogeneous glass
Radionuclides and hazardous species are atomically bonded in a durable glass structure usually to oxygen atoms that are also bonded to the matrix elements, Si, Al, В, P, etc., by short range order (SRO) and medium range order (MRO)	Some radionuclides and hazardous species are atomically bonded in a durable glass structure as with homogeneous glass, but other radionuclides reside in a very soluble immiscible glass phase (glass-in-glass phase separation)
Description
Radionuclide immobilization mechanism Key * + b Cs U Tc Pu xl.	Chemical incorporation	Chemical incorporation
Waste loading(s)/ durability	(a) moderate waste loading, (b) good overall durability, (c) easy to model radionuclide release from a single phase Joule Heated Melters (JHM), Advanced Joule Heated Melters (AJHM), Cold Crucible Induction Melters (ССІМ), Hot Isostatic Pressing (HIP), Hot Uniaxial Pressing (HUP)	(a) moderate waste loading, (b) poor durability for certain radionuclides, (c) impossible to model radionuclide release as the fraction of the second phase is dependent on thermal history Joule Heated Melters (JHM), Advanced Joule Heated Melters (AJHM), Cold Crucible Induction Melters (ССІМ), Hot Isostatic Pressing (HIP), Hot Uniaxial Pressing (HUP)
Immobilization technologies

Type of glass	Major structural components	Comments
(Si04) 4, (B04)-5, (B03)-3 and some (AI04) 5 and (Fe04)~5 structural units to which alkali, alkaline earth, and waste species bond.	Ease of processing, melt temperatures 1150-1200°C to minimize volatility; cold cap production if feasible minimizes volatility; most waste cations highly soluble in glass; overall waste solubility 25-40wt%; made by JHM, AJHM, induction melting, ССІМ or HIP.
Alkali borosilicate [3-8, 18, 34, 61]
Atomic structure of a French nuclear waste glass: unshaded region shows formation of a (Na, Cs)2Mo04 cluster [209].
Lanthanide borosilicate (LaBS) [5,6, 18, 210-214]	(Si04) 4, (B04)-5, (B03F3and some (AI04) 5 structural units to which lanthanides, alkaline earth, and other waste species bond.	Higher waste loading (16-59wt%) for actinides/lanthanides than alkali borosilicates; lanthanides serve as neutron absorbers; 1300-1500°C melting causes volatilization of some radio­nuclides; corrosion similar to alkali boro-silicates; made by ССІМ, HIP, or induction melting.
Atomic structure of a French HLW rare earth bearing borosilicate glass. Na+, Ca2+ and Nd3^ exist in the percolation channels. PR is the polymerized region and DR is the depolymerized region [7].

Type of glass Major structural components

(Si04r4 and (AI04)~5 structural units to which alkali, alkaline earth, and waste species bond (similar structure to borosilicate glasses when (В04Г5 are present).

Melt temperature of ~1600°C causes volatilization of radionuclides; waste loading dependent on rapid cooling, e. g. 20wt% U02 if cooled rapidly while <10wt% if cooled slowly; improved durability over borosilicate glass; ССІМ, HIP.

Atomic structure of a simple generic M203(G203)2 glass (M is modifying cations, G represents tetrahedral cations). The shaded regions are the PR regions. The un-shaded regions represent the percolation channels or DR regions (from [215]).

Aluminoborate (B04)“5, (B03)“3 and some (AI04) 5.

glasses

High silicate glasses (Si04)“4

(sintered glasses)

[4, 8, 61]

Alkali alumino — (P04)-3 and (AI04)-5 structural

phosphate [3-8, units to which alkali, alkaline

34, 217-220] earth, and waste species bond.

Requires hot pressing and sintering at 600-800°C in order to retain volatile fission elements such as Cs, Ru, Mo and Tc; waste solubility 5-35wt%.

Atomic structure of sodium silicate glass. Glass formers are small open circles, oxygen atoms are large open circles, modifier cations are small filled circles, U atoms which form clusters are large filled circles [216].

4

Atomic structure of phosphate glass with P4O10 cage-like structures which provide the basic building block for phosphate glass formers.

Melts at lower temperatures than silicate or borosilicate systems; most cations readily incorporated; accommodates >10wt% sulfate; corrosive to materials of construction; tendency to devitrify; durability comparable to borosilicate glass if alumina content is sufficient; composition ~ 24-27 Na20, 20-24, Al203 + MemOn, 50-52 P205; JHM, AJHM, ССІМ.

Type of glass Major structural components

Lead iron phosphate (LIP) [4, 8, 61, 221-227]

Atomic structure of LIP glass. Polyphosphate chains are cross — linked by lead atoms (open circles) and iron atoms (small filled circles) which form ‘knots’ in the percolation pathways that inhibit cation diffusion [228].

40-66 PbO; 30-55 P205; 0-10 Fe203 dependent on amount of iron oxide in waste; melts 850- 1050°C; waste loading (~20wt%); abandoned due to regulatory issues with PbO component poor solubility of certain species, devitrification, poorer waste solubility than borosilicate glasses, etc.; JHM, AJHM, ССІМ.

Iron phosphate (IP) [229-247]	(P04)-3 and (Fe04)~5 tetrahedral structural units to which alkali, alkaline earth, and waste species bond; Fe-O-P bonds have shown to be hydration resistant whether iron is Fe2+ or Fe3" [248]
Chalcogenide and chalcohalide [96, 251, 252]	Glasses obtained by melting chalcogen elements (S, Se, and Те) with Group V and IV elementsTe02-XCi-B203Te02- XCI-Li20Te02-XCi-Na20XCI = ‘mixed chlorides’ waste simulant at ~19wt%
Adapted from [11].

Atomic structure of IP glasses are nano-heterogeneous, with FeP04- like regions and phosphate chains that incorporate Fe2+/Fe3+ network­modifying cations. Large atom in center of cage like structure is a waste cation [249, 250].

Good chemical durability; high solubility for many heavy metals (U, Cr, Zr, Cs, Mo, noble metals, rare earths); melts 950-1100°C; viscosity typically <1 poise; low corrosion of oxide refractories and Inconel alloys; waste loadings 25-50wt%; tendency to devitrify; JHM, AJHM, ССІМ.

S, Se, and Те glasses for radionuclides difficult to immobilize in borosilicate glass systems, i. e. 12SI. Gels such as Pt2Ge4S96 are used to immobilize actinides, noble gases, carbon dioxide, and mixed chlorides.

[from 253]

Table 6.6 Examples of glass-ceramics and glass ceramic materials (GCM) as waste forms
Borosilicate based Alkali borosilicates [44, 45, 254-257]
Glasses which are allowed to partially crystallize in a stirred melt pool or upon cooling; crystals are inert; crystallized glass viscosity is non- Newtonian; secondary phases must be inert; AJHM or ССІМ. Electrorefiner wastes; radionuclide release from each phase is measured, e. g. Si, Al, Na, Li (sodalite and glass), В (glass), Cl, I (sodalite and halite); HIP or cold pressing/sintering. Zirconolite is major crystalline phase for Pu and Gd (neutron absorber); for low purity actinide wastes; Pu partitions into crystalline phase over the glass phase by a factor of 100:1; accommodates actinides and any associated impurities; HIP. Pyrochlore host for actinides and Sr; pollucite host for Cs and Bb; noble metal fission products form small metallic droplets. Melt temperatures from 1100-1400°C; controlled crystallization between 530 and 720°C; leaching characteristics have been noted to be comparable to the borosilicate glasses affording no significant advantages; additional work in this area has been limited, melt and control crystallization or press and sinter.
Borosilicate	NiFe204 spinels, ZrSi04, Al203
Glass bonded sodalites [49-51, 258] Synroc alumino — borosilicates [11]
Borosilicate 12SI in Nal, 12SI in sodalite, Na8AI6Si6024(l)2, Cl in sodalite, Na8AI6Si6024(CI)2 Alumino — Zirconolite, CaZrTi207 borosilicate
Barium aluminosilicates celsian [4, 8, 61, 259]	Borosilicate (sodium alumino-silicate with 2-7 wt% B203 and 3-4wt% Ti02)	Celsian, BaAI2Si208 pyrochlore (fttr2Ti207; RE-rare earth), Scheelite (BaMo04), Pollucite (CsAISi206) molybdenum-nosean [Na8AI6Mo04(Si04)6], Perovskite CaTi03, Diopside CaMgSi206, Eucryptite LiAISi206 spodumene LiAISi206, Nepheline, NaA1Si04

Waste loadings ~ 30wt% for European and Japanese commercial wastes which is usually ~16wt%; Melted at 1300°C; controlled crystallization in the range 800-1100°C; Cs was in the diopside; La, Ce, Nd, Pr in the perovskite, Sr and Sm were in the glass; noble metals were metallic.

Sphene and Synroc crystalline ceramic forms, mainly zirconolite, can also be formulated. Formation at 1300-1500°C. Actinides and BEEs, and Sr are in zirconolite; Cs and the remaining Sr into the vitreous phase; ССІМ and cool, press and sinter.

Formed by HIPing calcine (70wt%) with Si, Ті, Al metal and alkali oxides; for high Zr containing Idaho National Laboratory wastes.

Form at 1200°C. Fresnoite hosts Ba and Sr, priderite hosts Ba, pyrochlore hosts BE, actinides, BE and Sr. Cs remains in the glassy phase. Glass is 50% and crystalline phases are 50%.

Silicate based Basalt [4, 8, 61, 261. 272, 273]
Complex natural oxide based on Si, Ca, Mg, Fe, Al and Ті Alumino-silicate glass	For Purex wastes: augite (Ca, Mg, Fe)2Si206 powellite (Ca, Sr) Mo04 spinel (NiFe204).	Glasses melt in the range 1300-1400°C. Crystallization is carried out at temperature ranges 670-700°C and 900-950°C. The chemical durability superior to that of borosilicate glasses; JHM, ССІМ. Applications to commercial and defense wastes, including decontamination of Three Mile Island containment water together with core debris; melt at 1400-1500°C and controlled cooling after casting the glass into containers; JHM, ССІМ; arc melting. Cast glasses crystallized by holding at 1200°C for 16h; Ті phases retain the actinides; JHM, ССІМ; arc melting.
Iron enriched basalt (IEB) [4, 8, 61, 261]	Iron spinel; feldspars NaAISi308 to CaAI2Si208; augite (Ca, Mg, Fe)2Si206; fluroapatite, Ca5(P04)3F; zircon, ZrSi04; fluorite CaF2; cristobalite Si02; hematite, Fe203, mullite AI6Si2013 Same as above plus: zirconolite, pseudobrookite Fe2Ti05, chevkinite Ce4Fe2Ti3Si4022
Iron enriched basalt (IEB) with Ti02 and Zr02 [4, 8, 61, 261] Magnesium aluminosilicate (MAS) [4, 8] Phosphate based Apatite/monazite glass ceramics [274-276]
Alumino-silicate glass
Enstatite MgSi03lndialite/Corderite Mg2AI4Si5018	Used as an encapsulant for Zr alloy cladding wastes; accommodates 20% Zr02; press and sinter.
Magnesium alumino-silicate
Apatite Ca5(P04)3(F, CI)Monazite (Ce, U) P04	Apatite hosts Ca, P,F, Cl, S, Sr, Cs, As, Pb, Ba, Hg, Cd, Cr, U, and Ce, melted at 1400°C, crystallized at 1150°C and allowed to furnace cool; investigated primarily for phosphate-rich or fluoride-rich waste streams including Idaho National Laboratory CaF2 wastes; JHM, AJHM, ССІМ.
Calcium phosphate
Adapted from [11].

Table 6.7 Attributes of homogeneous and multiphase ceramic (mineral) waste forms

Waste form(s) Single phase oxides/ Multiphase oxides/minerals/

minerals/metals metals (granular or monolithic)

(granular or

monolithic)

(e)

high waste loadings superior overall durability difficult to model durability of species released from multiple phases and grain boundaries

need to tailor for species partitioning amongst phases

need to determine species partitioning and source terms from each phase may form an intergranular glassy phase that sequesters species of concern

may require precalcining for certain processes to work efficiently

Immobilization Hot Isostatic Pressing (HIPing >40 wt% crystals), Hot

technologies Uniaxial Pressing (HUPing >90 wt% crystals), Press

and sinter (> 90 wt% crystals), Fluidized Bed Steam Reforming (>90 wt% crystals)

Simple oxides X02 Oxides [277-280]

Cubic Zirconia, 1C

Murataite, 3C

Zr02, U02, Th02,Hf02, Pu02 have the simple fluorite CaF2 cubic structure; make by HIP, HUP, press and sinter, melt and crystallize.

Pyrochlore, 2C

Complex oxides Pyrochlore [281-284]

A derivative of the fluorite structure type, A2B207, where A-site contains large cations (Na, Ca, U, Th, Y and lanthanides) and the В-site contains smaller, higher valence cations (Nb, Та, Ті, Zr, Fe3+). Also a derivative of the isometric fluorite structure A6BuC5TXt0-x. with multiple units of the fluorite unit cell; hosts U, Pu, Cm, and BEs including Gd a neutron absorber. Forms in solid solution with pyrochlore. Monoclinic CaZrTi207 has a fluorite-derived structure closely related to pyrochlore, where Pu, U, Gd and Hf may be accom­modated on the Ca/Zr-sites, as in the case of Ca(Zr, Pu)Ti207.

V Zr4* • Y3+ • Tl4* • Zn2+ # Gd3* b Fe3* # 02~__________________

Murataite [3, 7, 285-293]

Projection along [111] direction for zirconia, pyrochlore, and murataite structures [from 284]

Zirconolite [294-299]

Crystalline ceramic phase Perovskite [296, 300]	Comments CaTi03 has a wide range of compositions as stable solid-solu­tions; orthorhombic; consists of a three-dimensional network of corner-sharing TiOe octahedra, with Ca occupying the large void spaces between the octahedra (the corner-sharing octahedra are located on the eight corners of a slightly distorted cube). Plutonium, other actinides, and rare-earth elements can occupy the Ca site in the structure, as in (Ca, Pu) Ti03. The octahedra can also tilt to accommodate larger cations in the Ca site [from 301].	Structure
[from 301]
Ba12(AI, Ti)8016 tunnels between Ti06 octahedra accommodate 133Ba, 137Cs and soSr.
Ba-Hollandite [302, 303]
Ferrite garnet [283] Garnet [304-307]
[slA3[6lB2[Ti04]3, e. g.[sl(Ca, Gd, actinides)[6lFe2[4lFe3012 A3B2(X04)3; distorted cubic structure; B06 octahedra and X04 tetrahedra establish a framework structure alternately sharing corners; A and В sites can host actinides, BEs, and X = Si4+, Fe3+, Al3+, Ga3+, Ge4+ and V5+ making silicate, ferrite, aluminate, gallate, germinate, and vanadate garnets.

Sr, La, Ce, Y positions are indicated by the solid circles. Other cations are in the octahedral positions, [from 309]

Freudenbergite [310]

[from 311]

Simple silicates Zircon/Thorite [312, 313]

ZrSiO/ThSiO,,; zircon is an extremely durable mineral that is commonly used for U/Pb age-dating, as high uranium concen­trations (up to 20,000 ppm) may be present; the PuSi04 end member is known and Ce, Hf and Gd have been found to substitute for Zr.

Crystalline ceramic phase Comments	Structure
Titanite (sphene) [261, 314] CaTiSi05 can sequester cations such as Ba, Sr, and fission product oxides (~15wt%), U, Cr, and Ni in the Ca sites (dark circles). Tetrahedra are Si and octahedral are Ті.
Britholite (silicate apatite) Also known as oxy — apatites in the literature. [54, 82, 87, 315-318]	(BEE, Ca)5(Si04,P04)3(0H, F); i. e. Ca2Nd8(Si04)602, Ca2La8(Si04)602; based on ionic radii of Nd3+, La3+, and Pu3+, an extensive range of solubility for Pu3+ substitution for the Nd or La, particularly on the 6h site, is expected. Since there is an extensive range in the Ca/BE ratio in these silicate apatites, a fair amount of Pu4+ substitution may be possible; La34 through Lu3+ can substitute for Ca2+ and form oxyapatites, ВЕ4і67іИо. зз[Зі04]зО; can also accommodate Sr and Cs, Th, U, Np.

Framework silicates Zeolites [320-325]

(Xxlnl(A2)x (Si02)y], where X is the charge balancing counter-ion, n is the charge of the counter-ion, x is the number of charge — deficient alumina sites, and у is the number of charge-neutral silica sites; characterized by internal voids, channels, pores, and/or cavities of well-defined size in the nanometer range, = 4-13A; channels and/or cavities may be occupied by charge — compensating ions and water molecules. Zeolites like Ag-Mor — denite selectively sorbs l2 (12SI); certain zeolites can be converted to condensed oxide ceramics by heating. This process is particularly attractive for waste form fabrication because capture and storage is performed with minimal steps.

Structure of Zeolite-A [326] showing alternate Al and Si atom ordering but omitting the tetrahedral oxygens around each Al and Si.

Pollucite [327-332] Pollucite Cs/Ti Version [333-342]

Crystalline ceramic phase Comments Structure

Nepheline [49, 343-348] NaAISi04 silica ‘stuffed derivative’ ring-type structure; some

polymorphs have large nine-fold cation cage sites while others have 12-fold cage-like voids that can hold large cations (Cs, K,

Ca). Natural nepheline structure accommodates Fe, Ті and Mg.

Leucite* KAISi206; К analogue of nepheline

Sodalite Na8CI2AI6Si6024 also written as(Na, K)6[AI6Si6024H2NaCI) to demonstrate that 2CI and associated Na atoms are in a cage structure defined by the aluminosilicate tetrahedra of six adjoing NaAISi04; a naturally occurring feldspathoid mineral; incorporate the alkali, alkaline earths, rare earth elements, halide fission products, and trace quantities of U and Pu (sodalite was and is being investigated as a durable host for the waste generated from electro-refining operations deployed for the reprocessing of metal fuel); minor phases in high level waste (HLW) supercalcine waste forms* where they retained Cs, Sr, and Mo, e. g. Na6[AI6Si6024](NaMo04)2; sodalite struc­tures are known to retain B, Ge, I, Br, and Be in the cage-like structures

Nosean, (Na, K)6[AI6Si6024](Na2S04)), silica ‘stuffed derivative’ sodalite cage-type structure host mineral for sulfate or sulfide species.

Hauyne, (Na)6[AI6Si6024]((Ca, Na)S04)1_2 sodalite family; can accommodate either Na2S04 or CaS04

Helvite (Mn4[Be3Si3012]S : Be can be substituted in place of Al and S2 in the cage structure along with Fe, Mn, and Zn

	Danalite (Fe4[Be3Si3012]S) Genthelvite (Zn4[Be3Si3012]S) Lazurite, (Ca, Na)6[AI6Si6024]((Ca, Na)S, S04,CI)x; can accommodate either S04 or S2,Ca or Na and Cl (Na, Ca, K)6[AI6Si6024]((Na, Ca, K)2C03)l 6»2.1H20 only found in hydroceramic waste forms
Cancrinite [358]
Crystalline SilicoTitanate (CST) [336, 359-363]
[(Ca, Na, K,Ba)AISi04 incorporates Ca, Na, K, Ba, Cs, and Sr
Crystal structure of Cs exchanged Nb-titanium silicate. The dark and light grey spheres represent Cs+ cations and water molecules, respectively [from 364].

The following dehydroxylated micas have been synthesized phase pure: ИА^ізО,,, ІМаАІзЗізОц, КАІзЗізОц, ИЬАІзЗізОц, СвАІзЗізОп, ТІАІзЗІзОц, Сао. БПо. БАЇзЗізОп, Зго. бПо. бАізЗІзОц, Вао. БПо. БАЇзЗізОп, Ьао. ззПо. ббАІзЗізОп. In the Cs-mica up to 30wt% Cs20 can be accommodated, in the Rb-mica up to 22wt% Rb20 can be accommodated, and in the Ra-mica up to 19wt% RaO can be accommodated. Mg, Fe2+, Fe3+, Mn, Li, Cr, Ті and V can substitute for Vl-fold coordinated Al3+.

Micas (dehydroxylated) [365-367]

[from 368]

Continued

Phosphates

Monazite [218, 274, 369-373] CeP04 or LaP04; very corrosion resistant and can incorporate a

large range of radionuclides including actinides and toxic metals into its structure; it has been proposed as a potential host phase for excess weapons plutonium and as a host phase for radionuclides and toxic metals in glass-ceramic waste forms for low-level and hazardous wastes.

Xenotime [218] YP04

Apatite [10, 54, 45, 218, 274, Ca4xRE6+x(Si04)6y(P04)y(0,F)2; actinide-host phases in HLW glass,

317, 318, 372, 375-384] glass-ceramic waste forms, ceramic waste forms and cement;

actinides can readily substitute for the rare earth elements in the crystal structure, as in Ca2(Nd, Cm, Pu)8(Si04)602, and fission products are also readily incorporated. However, the solubility for tetravalent Pu may be limited without other charge com­pensating substitutions; has been proposed as a potential host phase for Pu and high-level actinide wastes.

Sodium zirconium phos­phate (NZP) [218, 386-391]	NaZr2(P04)3; structure can incorporate a complex variety of cations, including plutonium; three-dimensional network of corner-sharing Zr06 octahedra and P04 tetrahedra in which plutonium can substitute for Zr, as in Na(Zr, Pu)2(P04)3. Com­plete substitution of Pu4+ for Zr has been demonstrated in NZP. Cs and Sr can substitute for Na while fission products and actinides substitute for Zr in octahedral positions. P is tetrahe­dral.
[from 388]
Thorium phosphate diphos­phate (TPD) [218, 392-394]	Th4(P04)4P207; a unique compound for the immobilization of plutonium and uranium; partial substitution of Pu for Th has been demonstrated (up to 0.4 mole fraction), complete substi­tution is not possible.
Continued

Aluminates

Magnetoplumbites [22, 62, Nominally X(AI, Fe)12019, where X = Sr, Ba,(Cs05 + La05) and (Na05 396-399] + La0.5>. The X site is XIl-fold coordinated and both Cs+/

Ba2+-Fe3+/Fe2+ or Cs /Ba2 — Ti47Ti3 — type substitutions can occur. Accommodating structures because they are composed of spinel blocks with both IV-fold and Vl-fold coordinated sites for multivalent cations and interspinel layers which have unusual V-fold sites for small cations. The interspinel layers also accommodate large cations of 1.15-1.84A, replacing oxygen in XIl-fold sites in the anion close packed structure. The large ions may be monovalent, divalent, or trivalent with balancing charge substitutions either in the interspinel layer (Na0i5 + La0i5) or between the interspinel layer and the spinel blocks (Cs+/ Ba2+-Fe37Fe2+ or Cs /Ba2 — Ti47Tr). [16]

Table 6.9 Attributes and examples of encapsulant and embedded waste forms (cements, geopolymers, ceramicrete, hydroceramics, and bitumen)
Waste form(s)	Encapsulated waste forms	Embedded waste forms
Liquid waste is mixed with concrete or other binder — hydrated phases occur that can incorporate the radionuclides or hazardous species weakly or retain them by sorption. Example Cs and U sequestered by C-S-H hydrates and U and Tc sequestered by secondary fly ash granules. The remaining species are trapped on the grain boundaries of the interlocking C-S-H phases. Encapsulation and some chemical incorporation	Liquid waste is mixed with concrete of other binder — primary phases and any secondary phases created by hydration (if an active mechanism) do not retain or sorb the radionuclide or hazardous species. Example shows Tc, Cs, U, and Pu all on the grain boundaries. Encapsulation and no chemical incorporation
Description
Radionuclide immobilization mechanism
Key О Phase or binder without radionuclides * $ I Cs U Tc Pu xl.
(a) low waste loadings, (b) lower overall durability, (c) difficult to model radionuclide release from hydrated secondary phases and grain boundaries, (d) easy to process — usually mix and cast	(a) low waste loadings, (b) lower overall durability, (c) difficult to model radionuclide release from grain boundaries, (d) easy to process — usually mix and cast, (e) in case of bitumen, must be heated to flow so embedding can occur
Waste loading(s)/durability
Mix and pour (cement, geopolymer, ceramicrete), Heat, mix, and pour (bitumen), Mix, pour, cure at slightly elevated temperatures (hydroceramics)
Immobilization technologies

Waste form(s) Composite	Composite
Multiphase granular oxides/minerals/metals (must be monolithed due to disposal requirements if not containerized)	Previously made waste forms in need of remediation (monolithing agents can be numerous and include glass — see GCMs above) Chemical incorporation and encapsulation/embedding
Description
Radionuclide immobilization mechanism Phase or О binder without radionuclides * # о — f % Cs U Tc Pu xl.
Chemical incorporation and encapsulation/embedding
(a) high waste loadings only if binder (monolithing agent) is minimized, (b) superior overall durability — double containment, (c) difficult to model radionuclide release from multiple phases, (d) need to tailor for and determine radionuclide partitioning amongst phases, (e) may require precalcining for certain processes to work efficiently	(a) high waste loadings only if binder (monolithing agent) is minimized, (b) superior overall durability — double containment, (c) difficult to model radionuclide release from multiple phases, (d) need to determine radionuclide partitioning amongst phases
Waste loading(s)/ durability
Immobilization technologies for matrix phase	Mix and pour (cement, geopolymer, ceramicrete), Heat and pour (glass or metal), Heat, mix, and pour (bitumen), Mix, pour, cure at slightly elevated temperatures (hydroceramic)

in simulated waste glasses (Table 6.4). These more highly ordered or polym­erized regions of MRO, often have atomic arrangements that approach those of crystals and are often referred to as quasi-crystalline species or quasi-crystals. Crystalline ceramics incorporate radionuclides and hazard­ous species by a combination of SRO, MRO, and LRO. The LRO defines the periodic structural units characteristic of crystalline ceramic structures. In glass, glass-ceramics, glass composite materials (GCMs), and crystalline ceramics, the radioactive and hazardous constituents are atomically bonded by a combination of SRO, MRO, and LRO. In GCMs there is additional encapsulation of the ceramic components in the glass matrix.

A maximum of 1% of SNF discharged from reactors could be defective. Volume expansion associated with the oxidation/hydration of the SNF matrix or zirconium may crack/unzip defective cladding (Cunnane et al., 2003). Figure 7.15 shows a schematic of this unzipping process (DOE, 2002). Unzipping was observed in the Argonne National Laboratory 1.5-year long tests, caused by stress generated by corrosion product accumulation in the gap of cladding and the fuel matrix from uniform corrosion of Zircaloy cladding at 175°C (347°F) (Cunnane et al., 2003).

Oxidation/hydration may occur with either residual moisture inside the intact canister or container, or from moisture that has intruded into the failed canister or container. This cladding failure may affect the magnitude of the radionuclide release fraction and challenge the retrievability of the SNF materials, and lead to configuration changes in internal structure that impact nuclear criticality.

(b)

As mentioned above, facilities for RAW storage (of the Radon type) are intended for the long-term storage of RAW containing short-lived radio­nuclides with a half-life of less than 30 years, including 137Cs and 90Sr. They only contain LLW and ILW. The RAW suppliers are the nuclear facilities of the nuclear industry (unconnected to the fuel cycle), organizations that operate nuclear reactors for research, and medical, training and scientific research centres that carry out radioisotope production.

The selection of suitable sites for RAW repositories was conducted firstly on the basis that any transfer of radionuclides into underground flowing water or its environs must be avoided. RAW repositories must therefore be placed in a clay massif, with low filtration and high sorption properties for radionuclides. The distance to the nearest water-bearing horizon must exceed 10 m. An area that met these requirements was identified 25 km from the town of Sergiev Posad in the Moscow region. From a geological point of view, the area offers sturdy layers of clay deposits of glacial origin (the Moscow and Dneprovsk moraines) that limit filtration and provide high sorption. The nearest water-bearing horizon within the limits of this area is located where the deposits of the Moscow and Dneprovsk glacial moraines meet at depth intervals of 38-42 m [5]. Groundwater in the area is not subject to regional propagation and remains local to the site, appearing only during autumn and winter in the areas adjacent to the man-made construc­tions. The filtration factors of the glacial clay deposits vary between 0.001 and 0.003 m/day, depending on the presence of sandy interlayers and the disturbance of the integrity of the base soil; these soils also have a high sorption capacity. Given that more than 90% of RAW contains 137Cs with a half-life of less than 30 years, 90Sr (29 years) and 60Co (5 years), the com­position of the soil makes the selected area an ideal location for RAW repositories, meaning that the radiation safety of the population outside the limits of the facility’s protection zone was guaranteed.

The construction of near-surface repositories for LLW and ILW began in the mid-1960s. A number of advances in repository design have since taken place, which have improved hermetic conditions and allowed the creation of reliable monitoring systems. Some key developments are:

• sunken monolithic repositories with a capacity of 400 itf [ a standard 1960s project designed by the USSR State Special Design Institute of Minsredmash (SSDI), a historical repository;

• sunken composite repositories, a historical type of repository designed by MosNPO ‘Radon’;

• sunken composite repositories with a ground-based tier, a 1980s devel­opment by MosNPO Radon;

• sunken monolithic repositories with a capacity of 5,000-10,000 m3 (SSDI);

• sunken monolithic repositories with a hangar superstructure, developed in the 1990s (SSDI);

• large diameter boreholes;

• drill-type repositories for SIS of the SSDI;

• repositories for SIS containing 226Ra, developed by MosNPO Radon.

Near-surface type historical repositories take the form of trench grooves 4-5 m deep, with the bottom of the trench covered with a hydro-insulating layer. The walls are made of reinforced concrete blocks or monolithic rein­forced concrete 0.4 m thick. The top of the repository is covered with reinforced concrete slabs and a layer of asphalt. Internally, the repository is made up of several sections. RAW in the repository was bulked in 1 m thick layers, which were then plugged with cement solution, prepared for use with LRAW with low salinity. In the mid-1980s an additional level, 3.5-4 m high, was built above some of the repositories, creating two-storey constructions with a capacity of about 20,000 m3.

Up until the 1990s these repositories were considered disposal facilities and were intended for the final disposal of LLW and ILW, with the surrounding rocks carrying out a basic barrier role to ensure geo-ecological safety. From the end of the 1990s, following IAEA recommendations, these near-surface type constructions were given a new status as ‘RAW repositories with a limited period of storage’. The new designation was based on the idea that adequate environmental protection from the hazards of RAW is dependent on man-made barriers, i. e. durable matrixes, RAW packing, backfill between packages, and the structural elements of the repository.

The SO (Fig. 11.3), which was intended to provide the environmental con­tainment of the damaged reactor, was erected between May and November 1986, under conditions of high radiation exposure of the personnel. The SO was constructed using steel beams and plates as structural elements. Its

foundation rests at some points on the original structural elements of unit 4, whose structural integrity is not well known. Thus the ability of the SO structure to withstand natural events such as earthquakes and tornados is not known with any certainty.

The SO has approximately 1,000 m2 of openings in its surface. These openings allow approximately 2,000 m3 per year of precipitation to perco­late through the radioactively contaminated debris. The collected water is contaminated with 137Cs, 90Sr and transuranic elements. The main potential hazard associated with the SO is a possible collapse of its top structures and the release of radioactive dust into the environment. Another concern related to the fuel-containing material is its possible transport out of the SO into groundwater through the accumulated water.

To avoid a collapse of the SO, some measures have been implemented to strengthen unstable parts of the SO and to extend their stability to 40 years. In addition, a new safe confinement (NSC) facility is planned to be built as a cover over the existing SO. The Ukrainian government supports the concept of a multifunctional facility with a service life of at least 100 years. This facility aims to reduce the probability of SO collapse, reduce the consequences of such collapse, improve nuclear safety, improve worker and environmental safety, and convert unit 4 into an environmentally safe site. The construction of the NSC is expected to allow the current SO to be dismantled and removal of fuel-containing materials from unit 4.

RAW management at the SSC ‘Complex’ and SSC ‘Technocentre’

SSC ‘Complex’ is responsible for the final disposal of waste at the RWDP ‘Buriakovka’, as well as for the monitoring of the RWDPs ‘Podlesny’, ‘The III line of ChNPP’, and multiple RWTSP (see Fig. 11.2). SSC ‘Complex’ also carries out the RAW collection and transportation within the ChEZ.

As the ‘Buriakovka’ facilities do not fully comply with the current requirements of surface disposal facilities, SSC ‘Technocentre’ started con­struction of the ‘Vector’ complex. At present, this new near-surface facility for low and intermediate level radioactive waste processing, storage and disposal, is under development. This complex will include:

• an engineering facility for the processing of all types of solid RAW;

• disposal facilities for short-lived solid RAW;

• storage facilities for long-lived waste (including fuel-containing materials);

• intermediate storage for vitrified HLW to be prepared for final disposal at a deep geological disposal facility.[27]