The management of fusion waste

Fusion Engineering and Design 14 (1991) 37-47 North-Holland

37

The management of fusion waste R. Hancox and G.J. Butterworth AEA

Fusion,

Culham

Loboratory

(Euratom/UKAEA

Fusion

Association),

Abingdon,

Oxon

OX14

,308,

United

Kingdom

Submitted 3 September 1990; accepted 6 November 1990 Handling Editor: P. Komarek

Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET; and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal.

1. Introduction Fusion power, based on the nuclear fusion of light elements to yield a net gain of energy, is being studied as a potential addition to fission power as a long term

energy source. The attraction of fusion power is that it usesfuels which are abundant and that neither the fuels nor most of the reaction products are radioactive. Thus fusion power has the potential to extend the world’s energy resources in a way which is environmentally attractive. In common with fission reactors, fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste which results from the neutron irradiation of the structural materials and the absorption of the tritium fuel. An important issue in fusion research, therefore, is whether the volume and hazard potential of this radioactive waste can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. In the case of power reactors, there are several possible approaches to the waste problem. The most attractive would be the use of alternative fuel cycles such as D-3He which avoid the use of tritium and result in the generation of fewer neutrons, but since the plasma confinement requirements for these fuel cycles are much harder to achieve it is unlikely that they could be used for the first generation of power reactors. The second possibility is the use of ‘low activation’ structElsevier Science Publishers B.V. (North-Holland)

ural materials in which significantly lower levels of radioactivity are induced; such materials were reviewed in an invited paper [l] at the last Symposium on Fusion Technology. A third possibility, connected with the use of low activation materials, is the recycling of structural materials, ie fabricating new reactor components from materials recovered from used components. A review [2] of the economic and environmental benefits of recycling steels from a fusion reactor indicated some advantages when the scale of the process is large enough. An essential approach to reducing the volume and environmental impact of fusion waste is through reactor design. The quantity of material required in reactor construction can be reduced by adopting compact designs, whilst the volumes and activities of waste streams can be reduced by designing for lower activation and by preventing neutron streaming through penetrations. It should also be possible to reduce the cost of waste disposal by the choice of appropriate cost-effective means of disposal. The hazards resulting from fusion waste are in many cases substantially less than from fission waste, and the revision of regulations to take this factor into account would lead to a significant reduction of the cost of disposal for fusion waste. The above comments all relate to power reactors, which may be in operation in the second half of the next century. In the shorter term, fusion experiments such as the Joint European Torus (JET) and experimental reactors such as the Next European Torus (NET) will generate radioactive wastes. For the experimental

38

R. Hancox, G.J. Buttenvortll / The managemettt of /usion waste

reactors, many of the above approaches to the reduction of the volume and environmental impact of the waste are inappropriate, since the total volumes involved will be too small to justify the substantial developments required or the timescale too short to allow significant changes. However, the appropriate design of such machines will be important, both because of the impact that design will have on the volumes of waste from the machines themselves, and because of the lessons to be learnt for subsequent power reactors. In the following sections the present understanding of the waste from JET, NET and power reactors is summarized. In the case of the waste from power reactors, the volumes and the environmental hazards are compared with those for fission reactors.

2. Waste from JET

JET is the first fusion experiment which produces significant numbers of neutrons and which will generate radioactive waste requiring disposal. In operation with deuterium, JET has already produced 3.4 X lOI6 neutrons/s from D-D reactions (equivalent to 100 kW of fusion power), and it is estimated that during full power operation with deuterium-tritium it will produce 10” neutrons per pulse from D-T reactions. The maximum credible number of neutrons which will be generated by JET, which has been used as the basis of some of the results quoted in this paper, is 10z4 neutrons; but a more realistic basis for planning is now thought to be 1O23neutrons. According to the present JET Statutes, the United Kingdom Atomic Energy Authority as the Host Organization is responsible for decommissioning JET and disposing of the waste. These requirements have been considered in a Preliminary Decommissioning Plan [3], and in particular the quantities and levels of activity of the decommissioning waste have been estimated, packaging volumes proposed, and the cost of waste handling and disposal considered. Whilst there will also be some radioactive operating waste, this will be relatively small in volume compared with the decommissioning waste, and is not considered in this paper. The decommissioning of JET and the disposal of the resulting waste are of both local and Community importance. The original planning permission for the construction of JET at Culham required that the plant and buildings be removed as soon as the project was complete. However, a change in NIREX strategy in 1987 recommending that decommissioning waste should go to deep disposal instead of shallow disposal has meant

that the planning permission has had to be extended until the year 2015, by which time a deep repository should be in operation in the UK. This change of policy, together with the international agreement in 1983 to stop the disposal of radioactive waste at sea, is expected to substantially increase the cost of waste disposal. The increasing cost of disposing of radioactive waste from JET is also important for subsequent fusion experiments such as NET and fusion reactors. Although the regulations for radioactive waste disposal vary at present between the member states of the Community, it is probable that deep disposal will become the normal requirement. Thus experience with JET will give guidance for other fusion devices, and the design of these devices should take account of the high costs of disposal. Experience in decommissioning and remotely dismantling JET will also be relevant to the design of these devices, but falls outside the scope of this paper. 2.1. Estimated

levels of activity

The levels of neutron induced activity of the components of the JET machine, the auxiliary heating equipment and diagnostics and the torus hall have been calculated using the FISPACT code and the UKACT 1 library. The calculations assumed that 1O24 neutrons would be generated over a 2 year period of full power D-T operation and since rate effects are insignificant, the levels of activity are proportional to the assumed number of neutrons. The specific activities of several machine components are shown in table 1. Where the activity level for any component has been calculated for several regions the activity level shown is the highest value, since this will usually determine the waste category. Based on these activation levels and the masses of each component, it is estimated that the total neutron induced activity of the machine 2 years after shutdown will be 1300 TBq. The levels of heat generation after shut-down in components of the JET machine have also been calculated, but are only significant in the vacuum vessel. Even in this case the heating falls below 1 W/te within 2 years, and therefore will not be a limitation on packaging and storing decommissioning waste. The expected limitation in a deep repository is an average heat generation of 2 W per 500 litre drum, corresponding to 4 W/te at a packing density of 1 te/m3. In addition to neutron induced activity, the vacuum vessel and in-vessel components will be radioactive as the result of absorbed tritium. The concentration of

R. Hancox, G.J. Bullet-worth / The management o/fusion waste

that the total activity in the torus hall will be 0.07 TBq at 10 years after shutdown. The highest dose equivalent rate at the surface of the torus hall concrete 20 years after shutdown is 3.2 /.&v/h. For a worker employed on 8-hour shifts for 250 days per year the upper limit of the possible annual hands-on dose equivalent is therefore 6.4 mSv, which is below the UKAEA target figure of 15 mSv for the maximum dose to an individual. In this case hands-on decommissioning seems possible, subject to better definition of the relevant dose rates and to possible restrictions arising from the hazards of inhaling or ingesting active demolition dust.

Table 1 Specific activity of JET components (GBq/te) Time after shutdown 2 years 6.4E+ 3 * Vacuum vessel Vacuum vessel equipment Graphite tiles 1.3E-2 Beryllium tiles 9.8 Toroidal field coils l.OE+2 * Copper Steel 1.4E+3 * External mechanical structure Inner wall 1.9E+3 * Outer wzill 8.8E+l + Poloidal field coils (Pl) Copper 1.3E-1 Steel 3.1 Transformer core Limbs 1.9 Centre piece 9.9

15 years 2.1E+2 * l.OE-3 3.1E-1 4.1E+ 1 * 6.OE+ 1 * 6.6E+l 3.5

*

2.2. Quantities

3.6E-2 1.2E-1 9.4E-2 4.OE-1

Components marked with * are ILW.

tritium has been estimated both on the basis of the basic absorption mechanisms and from preliminary measurements

during

D-D

operation.

Assuming

39

gra-

phite tiles, it is estimated that 30 g (11000 TBq) of tritium could be absorbed but that this could be reduced to 3 g (1100 TBq) by bake-out and glow discharge cleaning procedures. With beryllium tiles the tritium levels may be a factor 10 lower. The diagnostics and auxiliary equipment on JET consist of a large number of items of various sizes distributed throughout the torus hall and constructed from a wide range of materials. Therefore, it is not practical to calculate the activity of each diagnostic and piece of auxiliary equipment at this stage. An alternative approach has been adopted, in which the activity of a range of materials has been calculated on the basis of a unit neutron dose and a small range of neutron spectra, so that the activity of any component can be estimated when its composition is specified together with the neutron flux and spectrum to which it is exposed. For waste disposal purposes the quantities of the most important materials in various regions of the torus hall can be added together to obtain a first approximation to the total quantities of Intermediate and Low Level Waste. The activity of the internal surface of the torus hall has been calculated at various locations, and the activity as a function of depth into the wall has been calculated at two locations. From this information it is estimated

of waste

The quantities of radioactive decommissioning waste from JET have been estimated on the basis of the current UK definitions of waste categories. Since only p and y activity is involved, these categories allow the waste to be dealt with as follows: Material with an activity of 0.4 MBq/te or less, defined as Very Low Level Waste (VLLW), could be buried on site. Material of activity less than about 4 MBq/te could be buried in a local tip at a depth of at least 1.5 m. Material of activity less than 12 GBq/te (including any VLLW which cannot be disposed of locally) will be treated as Low Level Waste (LLW) and will be disposed of in a NIREX repository. Intermediate Level Waste (ILW), with an activity above 12 GBq/te, will be disposed of in a NIREX repository. Two packaging schemes have been considered. In one scheme no weight limit was imposed on individual packages, and in the other a weight limit of 25 te was imposed. The results of the latter scheme are quoted in this paper. Each machine component was considered, and a means of cutting it devised so that it would fit into the appropriate package - a 2.75 m3 steel box for ILW and a 5.4 m3 box for LLW. In the case of the diagnostic and auxiliary equipment, the active gas handling equipment and the torus hall concrete a simpler approach has been adopted based on an average packing fraction. The resulting volumes of packaged waste are shown in table 2 on the assumptions of 1O24neutrons, 2 or 15 years delay before decommissioning and waste packaging, and a 25 te limit on packages. It will be seen that the largest component of the JET waste is that part of the concrete from the torus hall (- 25%) which must be disposed of as LLW. It may be noted that if all the material from the wall, roof and

40

R. Hancox,

G.J. Buttenvorth

Table 2 Volumesof packagedwastefrom JET (m3) Immediate Delayed decommissioning (ie 2 years) (ie 15 years) ILW requiringshielding Machine Diagnosticsand auxiliaryequipment ILW not requiringshielding Machine Diagnosticsand auxiliaryequipment Activegashandling system Operationalwaste LLW Machine Diagnosticsand auxiliaryequipment Activegashandling equipment TorusHall Operationalwaste

220

0

71

0

438

542

142

11

125 1

125 1

997

679

2099

2260

577

861

1740 16800 209

1740 13800 209

21425

18870

floor (36,700 te) were well mixed the average induced activity would be 1.9 MBq/te 10 years after shutdown. This may be compared with the natural activity in the coal ash from a coal fired power station such as Didcot, which produces about 53,000 te of ash each month, with a specific activity of 2.6 MBq/te due to long-lived uranium and thorium isotopes. The biological hazard potential due to inhalation of the concrete and coal ash have been estimated to be 7 X 10m6 and 4 x 10v4 %/kg respectively. The cost of disposal of the waste from JET will therefore be dependent on the possibilities of local shallow disposal of material with very low levels of radioactivity. 2.3. Tritium in JET waste Tritium in the form of HT or HTO is a mobile radio-nuclide and can diffuse readily through many common materials including plastics and concrete. The presence of tritium in some JET components therefore imposes additional constraints on the storage, transport and disposal of waste.

/ The management

offusion

waste

All waste will remain for a period on the JET site prior to disposal, and the escape of tritium from stored wastes may be expected. Tritium decay also results in ‘He formation which, in addition to other radiolytic off-gassing, will pressurize any sealed waste container. Two courses of action are possible; either to keep the waste in an unconfined state and remove escaping gases with a ventilation system, or to store the wastes in sealed packages and minimise gaseous releases. Aerial discharges from the JET facility during decommissioning must comply with the site Authorization which has yet to be negotiated, and if tritium releases in storage were suitably low then unrestricted discharge to atmosphere could be acceptable. Alternatively, the air detritiation system to be installed as part of the Active Gas Handling System could be used to control aerial discharges, albeit producing an additional secondary waste (tritiated water). Adoption of the second option would require the use of a high integrity package, such as a welded steel drum, which should also meet the necessary criteria for transport and final disposal in order to avoid repackaging of the waste. If such special requirements were identified for JET decommissioning waste, they could involve additional costs for the development and authorization of suitable designs. The presence of tritium in waste sent for deep disposal may require that it be placed in a separate section of the repository or that additional ventilation be provided. It is not yet possible to be more specific, nor to estimate the implications of any such requirements on the cost of disposal. In the extreme situation it may be necessaryto detritiate the waste before packaging, which would add considerably to the costs of disposal. 2.4. Organics in JET waste Another problem associated with JET waste is the presence of significant proportions of organic materials in some components. Studies of the effects of organic materials on repository performance have shown that the solubility of some radionuclides may be enhanced and their sorption reduced. Since organic materials are used in JET in the field coils, transformer core and diagnostics, it is necessaryto consider the general problems of the disposal of organics and possible means of separating them from metallic waste. If, in due course, the effects of organic materials are proved unacceptable in a deep repository, the maximum requirement will be for a process to destroy the organics. Two methods for the segregation of organics have been proposed; chemically and mechanically. A further possible method involves the melting of coil assembliesin

R. Hancox,

G.J. Butterworth

an induction furnace to produce copper ingots, with glass fibre forming a slag and the resin being gassedoff. As with the presence of tritium, these processeswould require the use of specialized equipment and represent an additional cost.

3. NET

wastes

Collaborative studies involving the NET Team and several European laboratories are in progress to quantify the radioactive wastes to be expected from NET and to develop suitable management strategies. The most appropriate treatments for the structural waste prior to disposal will depend on the disposal site chosen and on the national regulations governing waste treatment and disposal in the host country of NET. At present there existsa wide variation within the European Community in the approach to waste management, and therefore it is necessary to consider the situation in several countries. 3.1. Waste inventory

Current studies on management of NET wastes assume an operating regime comprising a Physics Phase, in which a first wall fluence of about 0.03 MWy/m’ is anticipated, followed by a Technology Phase with a fluence of 0.8 MWy/m’. Since neither the design nor the operational mode of the reactor is yet fixed the fluence limits, activity levels and quantities of waste should be regarded as initial estimates only. The mass and activity of the principal structural components at various cooling times are shown in table 3, corresponding to an integrated power loading of 0.8 MWy/m’. The inventories are derived from the NET shielding

Table 3 Massesand activitiesof NET components Mass W Shieldingblanket 695 Inner shield 1018 Outershield 2218 Toroidal field coils Steel 3200 Other material 1400 Diverters(CaseA) TZM 14 MoRe 4

Activity (Bq) at 3SE18 4.9E17 3.2E17 4.4E16 2.1E16 3.2E15

1ooy 6.8E15 l.OE15 1.2E14

7.4E12 2.6E12

1.3ElO 1.3ElO

1Y

1OY

9.6Ell 2.6Ell

l.OE14 9.7E13 1.9E14 4.5E13

1.2E14 4.9E13

/ The management

of fusion

41

waste

blanket design, with predictions of the radionuclide content by JRC Ispra 141. Additional operational wastes will arise from replaced components arising, for example, from the testing of alternative blanket segments and from coolant, vacuum and fuel processing systems. Wastes associated with component replacement are expected to be similar in activity and composition to structural wastes from machine decommissioning, and their mass is likely to represent no more than 10% of the decommissioning waste for each complete replacement of first wall, blanket and back plate. 3.2. Waste management

strategy

An outline route for handling and disposal of NET wastes has been considered. After dismantling of the machine, short-term storage on site will be needed for reduction of decay heat and dose rates of the most highly activated parts. Large components will then require size reduction and sorting according to material, reuse value, activity and tritium content. Materials contaminated with tritium may need to be detritiated, using processesspecific to the material. For metals, melting offers a means of removing tritium to a level below 400 GBq/te as well as minimising volume [5]. Following any conditioning that might be needed, e.g. tritium immobilisation, the waste will be placed in interim storage pending diminution of the decay power to negligible levels. Storage times could vary from a few years to several decades, depending on composition. At the end of NET operations most of the components listed in table 3 fall within the category of Intermediate Level, or in some casesHeat-Generating waste, the exception being the toroidal field coil materiais which fall into the Low Level class (UK criteria). After 50 y storage the toroidal coil materials may reach the LLW lower limit and could be considered for nearsurface disposal or recycling for reuse in the nuclear industry, e.g. as waste containers or inner shielding for these containers. If the trend toward geological disposal of both ILW and LLW continues then all the other components will need to be despatched to a geological repository. 3.3. Disposal assuming Swedish criteria

Sweden and Germany have relatively advanced plans for management of radioactive waste. In Sweden, the first repository SFR 7, for LLW and short-lived ILW, opened in 1988. It is located at a depth of 50 to 60 m in granitic rock beneath the Baltic sea. In the same vicinity

42

R. Hancox,

G.J. Burrenvorth

Table 4 Volumes of NET waste

Swedish Cooling time(y) Shieldingblanket 50 Inner shield 50 Outer shield 20 Toroidal field coils 1 Steel Other material 1 Divertors(CaseA) TZM 10 MoRe

Volume (m3) 481 706 857

German Cooling time(y) 3 50 50

Volume (m3) 111 702 504

1236 541

1

1

520 228

17 8

1

29

1

12

10

a repository SFR 3 for decommissioning waste is planned for the year 2010. Deeper repositories SFL 2-5 at a depth of 500 m are in the design stage, with operation foreseen after 2020, for heat generating waste (HGW) and long-lived ILW. In Germany, plans are being developed for two repositories. The Konrad disused iron ore mine at a depth 800 to 1300 m will accommodate non-heat generating wastes from about 1991, while the Gorleben salt dome repository is expected to be suitable for all types of waste, with operation starting in the period 2000 to 2010. The volumes of packaged NET wastes based on Swedish and German disposal rules have been estimated by the Studsvik group [6], and are shown in table 4. Of the structural components only the outer shield, toroidal field coils, and LLW or ILW from the processing systems could be consigned to SFR. The more active near-plasma components would need at least 1 y of cooling before transportation could be undertaken. After 30 to 50 y intermediate storage, disposal of the shielded packages should be possible in SFL. Assuming detritiation and volume reduction by melting and casting, the 7000 te total mass of the steel components would give rise to a packaged volume of about 4000 m3, including the toroidal field coils. Of this total, 70% can be consigned to a repository of the SFR type and 30% would require the greater isolation of the SFL type repository. Each replacement of first wall, blanket and back plate during the lifetime of NET would generate 400 m3 of additional packaged waste.

/ The managetneru

Disposal

assuming

German

waste

be consigned to Gorleben, whilst inner and outer shields and divertor (caseA) could be disposed in Konrad after 50 y intermediate storage. In this scenario, a minimum of 2000 m3 of packaged waste, including the toroidal field coils, would require disposal. Restrictions on tritium content and outgassing rate may be important for fusion wastes. The limitation of 470 to 960 GBq/package for Konrad may be compared with a residual tritium concentration of less than 400 GBq/te found in melted or heat treated steel, which would permit packages containing 1 to 25 te of steel waste. The restriction on tritium outgassing of about 1.5 GBq/y per package may be compared with a rate of about 0.15 GBq/y-te observed in experiments at CEA. Thus it may be possible to accommodate up to 10 te of tritiated steel waste in a single package without special gas-tight containment. Tritium outgassing rate limits proposed for Gorleben lie in the range 1.1 to 110 MBq/y per package for disposal in tunnels and 190 to 1900 MBq/y per package for bore holes. Thus tunnel disposal of packages containing 0.1 to 10 te or bore hole disposal with 1 to 10 te of detritiated steel waste per package should be possible without gas-tight sealing. The limitations on radionuclide content and tritium release rates of packages according to Swedish disposal rules are not yet defined. 4. Power reactors Fusion power reactors will generate both operating and decommissioning waste. At the present state of development, very little is known about the operational wastes except the structural and breeder materials resulting from routine replacement of the blanket and divertor plates which are conveniently included with the decommissioning waste. The quantities of operational wastes from tritium processing are expected to be small compared with the waste from fuel reprocessing for fission reactors. For decommissioning waste, the two main issuesare the volume of waste and its hazard potential. The former is only important from the point of view of the cost of storage, transport and disposal Of greater importance are the hazard resulting from waste processing and the environmental impact of waste after disposal. 4.1.

3.4.

of fusion

Volumes

of decommissioning

waste

criteria

If the NET structural wastes were to be disposed of in the German repositories the shielding blanket could

The volume of raw and packaged waste from fusion reactors was studied at Culham on the basis of six specific D-T reactor designs - three tokamaks, two

R. Hancox, Table 5 Power reactor

Fusion

waste

(m3/GWe)

Non-active waste

Repositorywaste (Packaged) Wet)

1720 2840

1940 1990 8050

reactors

STARFIRE DEMO R7.54 PCSR-E CRFPR TITAN WITAMIR EEF-R2 Fission

G.J. Buttenvorth

1940 0 0

910 1120

480 2590

3900 4000

10400 10700 43400 4870 6030 21000 18000

reactors

PWR- without fuel reprocessing - with fuel reprocessing

13000 25000

reversed field pinches and a tandem mirror. These reactors cover a range of characteristics - for example the tokamak reactors include shielding between the blanket and superconducting coils, whereas the reversed field pinches have less shielding and normal coils. In all cases,material from replacement blankets and divertors was included in the total. The volume of waste from the Reference Reactor (EEF-R2) for the Study Group on the Environmental, Safety-related and Economic Potential of Fusion Power [7] (EEF) has also been calculated on the same basis by Sowerby and Forrest [8]. The totals of packaged waste based on the current UK definitions of waste categories are shown in table 5, and range from around 5000 to 43000 m3/GWe (the volume quoted for the EEF-R2 reactor excludes contaminated water). The lowest volumes were from the reversed field pinches, reflecting the fact that they are more compact reactors than the tokamaks or mirror. The PCSR-E reactor gives the largest volume for several reasons, including the fact that all material from the shield is assumed to be repository waste whereas for Starfire and Demo a substantial proportion of the shield is non-repository waste. This emphasizes the point that the shield contributes the largest component of the waste, and suggeststhat the use of low activation steels and suitable design may have a substantial influence on the volumes of waste produced. Extensions of the shielding around penetrations for divertors, vacuum pumping, auxiliary heating systems, etc, are often not well quantified in present studies, but could contribute significantly to the waste produced. In the USA a Senior Committee on Environmental, Safety and Economic Aspects of Magnetic Fusion Energy [9] (ESECOM) have estimated the radioactive waste

/ The management

o//usion

waste

43

produced by ten tokamak and reversed field pinch reactors, including D-T and D-3He fuelled reactors and hybrid fission-fusion reactors. The unpackaged waste volumes quoted for six of these reactors using D-T fuel and conventional electrical generation plant fall within the range 480 to 3360 m3/GWe, the lowest value corresponding to a compact reversed field pinch and most of the high values to tokamaks. The volumes of unpackaged waste deduced in the Culham and the ESECOM studies have been compared as functions of the mass power density, defined as the power output per unit mass of the fusion core. The two sets of volumes are comparable, and the trend to lower waste volumes for designs with a high mass power density is noticeable. This would be expected since the shield of a reactor is the major component of the waste volume and is also a major component of the mass of the fusion power core. The scatter in results in the two studies suggests that the volume of waste from a fusion reactor is not known within an accuracy of a factor 3. A mass power density of greater than 100 kWe/te has been proposed as a criterion for the economic viability of a fusion reactor and, if this is accepted and the shield design is optimised so that a significant proportion of the waste does not require disposal in a repository, the volume of packaged waste from a tokamak power reactor should be in the range 7500 to 15000 m3/GWe. The Culham estimates of the volume of packaged waste from a fusion reactor may be compared with the waste from the operation and decommissioning of a PWR reactor, which is also shown in table 5. In making the comparison it should be noted that the figures for the fusion reactors only include the reactor core itself and exclude externals such as heat exchangers, secondary heat transfer system, diagnostics, tritium processing plant or contaminated remote handling equipment, and the figures quoted for the fission reactors have been adjusted to compensate for these differences. The ESECOM report also presents a comparison of the volumes of waste from several fusion reactors and a LWR fission reactor without reprocessing. The volumes of waste from the activated structure of the fusion reactors (23 to 190 m3/GWe) are about twice the volume from a fission reactor (14 to 115 m3/GWe), but the larger volumes of low level waste from the rest of the plant (400 to 1100 m3/GWe) are comparable. The EEF Study Group makes a similar comparison between the packaged waste from the fusion Reference Reactor (14200 m3/GWe) with the waste from stage 3 of decommissioning of a PWR (1970 m3 and 670 te of spent

44

R. Hancax, G.J. Buttenvorth / The management of fusion bvaste

fuel/GWe), on the assumption that the large volume of waste from stage 1 and 2 decommissioning of the PWR (12 000 m3/GWe) is similar to the additional volume of waste from outside the core of a fusion reactor. The general conclusions of such comparisons are that although the waste from the core of a fusion reactor may be a few times greater than the corresponding waste from a modem fission reactor, both volumes are less than the volumes of waste from external components and less than the volumes of waste from fission fuel reprocessing plants. Thus the cost of waste disposal for fusion, if primarily based on the waste volume, will not significantly exceed the cost of waste disposal for fission reactors. An important lesson for fusion is the need to design the complete power station to limit the quantities of radioactive waste from outside the reactor core, so that this does not unduly augment the unavoidable waste from the core itself. 4.2. Environmental

impact of reactor waste

Whilst the major contribution to the volume of waste from a fusion power plant may be the shielding and equipment outside the reactor core, the major contribution to the environmental impact of fusion waste is almost certainly from the highly irradiated blanket and divertor components close to the reacting plasma. The simplest measures of the hazards due to irradiated components are their levels of activity or the surface dose rates of unshielded material. By these criteria, fission and fusion waste are similar in the early years after shutdown of the reactors. However, these are not adequate measuresof the long-term risks resulting from the storage and disposal of the waste. Better, but still incomplete, measures of the hazards are the Biological Hazard Potentials (BHP) due to inhalation or ingestion, which take account of the health effects of specific isotopes. For short times (- 1 year) after shutdown the waste from fusion has a hazard potential which is one or two orders of magnitude less than that for fission waste, and for larger times a substantially greater advantage. The environmental impact of waste following its disposal in a repository can only be fully assessed through a systematic modelling of the pathways by which radioactivity can return to the human environment. Such analysis should take into account both the normal development of the repository, involving the slow leaching of the waste by ground-water and the transport by water back to the surface, as well as intrusion into the repository by accident either through

use of the site or drilling. Such methods of analysis have been developed for the safety assessment of proposed repositories for fission waste, but are not easily applied to fusion waste because of the lack of nuclear and transport data for some fusion-specific radionuclides; in neither case can there be significant operating experience because of the long timescales involved. For shallow land burial, which is considered as an option for waste disposal in the USA, risks due to intrusion exceed those resulting from normal development of the repository. The ESECOM study calculated two intrusion-based indices for the waste from various fusion reactors and compared them with the waste from a fast breeder reactor (LSPB). The intruder dose for each reactor component is the maximum dose during the period 100 years after shutdown (when control of the site has lapsed) to 1000 years to a person who either builds a house or lives on the site and consumes vegetables, meat and milk produced on the site. This index does not take into account the volume of waste, and must not exceed 5 mSv/y. The annualized intruder hazard potential is the sum of the products of the intruder doses and the annual waste volumes. This annualized intruder hazard potential varies by a factor 1000 between the seven fusion reactors considered but for several cases is a factor of 100,000 less than the value for the breeder reactor and even the worst case is a factor 200 less. A preliminary assessment of the risks from fusion waste has been made by Smith and Butterworth [lo] for both shallow and deep burial, and considering normal development and intrusion. For this study only the first wall and blanket structural material was considered, since this is known to give the largest contribution to the risk, but three materials were considered which were typical of a standard ferritic steel (FV448), a low activation steel (LA12TaLC) and a lower activation vanadium alloy (V-3Ti-Si). In each case possible impurity elements were included, and it was assumed that the repositories contained 100,000 tonnes of the fusion waste. The maximum individual doses for the low activation steel are shown in table 6. The first conclusion of the study was that for all three materials the estimated risks for a shallow engineered repository were above the UK requirement which states that “the appropriate target applicable to a single repositoty at any time is . . . a risk to the individual in a year equivalent to that associated with a dose of 0.1 mSv; about one chance in a million” (based on a probability of developing fatal

cancer of 0.012 per Sv). However, in the caseof the two low activation materials the estimated risk did not exceed the requirement by more than a factor 5, which

R. Hancox,

G.J. Butterworth

suggests that all other components of a fusion reactor might be suitable for shallow burial if this option was available. The study also concluded that the risks associated with deep disposal were below the UK requirement and, in the case of the two low activation materials by more than a factor 300. A similar estimate of individual and collective doses to the public resulting from the release of radioactive materials from a geological repository has been published by Aggeryd and Bergstrom [ll]. A multi-compartment model was designed to describe the biosphere, with 38 nuclides passing to a well or lake. The doses resulting from drinking the water and from consuming fish, meat, milk, vegetables or cereals from the adjacent land were calculated. The maximum individual dose rates to the critical groups resulting from the release from the repository of 1 Bq/y of some important nuclides are shown in table 7. In determining the total dose rates these values must be multiplied by the individual release rates, which can vary greatly between elements. Because no account is taken of the delay in the release of radioactivity from the repository, the times at which the maximum dose rate occurs for some individual nuclides are much shorter than those predicted in other assessments. The most comprehensive assessmentof the doses to the public due to fusion waste in a deep repository has been made by Sowerby and Forrest as a contribution to the EEF Study Group, and a comparison made with

Table 6 Maximum individual dose for disposal of lost of LA12TaLC in a deep repository (Sv y-‘) Element C

N Si Fe Ta W Impurities Al

A8 Nb Bi Ir MO Total dose @v/y) Total risk (y)

Normal evolution

Intrusion

1.9E- 14 4.6E- 15 5.4E- 11

5.4E-9 3.4E-6 l.lE-7 2.3E-6 6.9E-6 1.3E-4

1.8E- 10 2.3E-7 3.3E-10

6.OE-5 1.6E-5 7.8E-6 4.5E-6 1.4E- 10 2.3E-7 2.9E-9

2.4E-4 1.4E- 10

/ The management

offusion

Table I Maximum individual

c-14 Mn-53 Fe-55 Co-60 Ni-59 Ni-63 MO-93 Ag-108m Eu-154 Re-186m PO-209

waste

45

dose rates from annual releases of 1 Bq Maximum individual W/ye=) 2.OE- 14 6.8E- 14 3.9E-15 1.6E-13 1.5E-15 4.OE- 15 l.OE- 14 1.5E-13 5.2E- 14 3.8E- 14 1.5E- 11

Time for maximum (years) SE+1 5E+4 lE+l 4E+1 4E+l 4E+l 2E+l 2E+2 5E+l 7E+2 2E+2

waste from fission reactors. The assessment assumed that transport of radioactivity by ground-water was the major hazard, rather than the release of gasesor intrusion. The near-field model took into account containment provided by metal canisters and the chemical environment existing within the cement-based repository (these were omitted in the Smith and Butterworth estimates). Water is then assumed to eventually enter a stream, which is used as the sole supply of drinking water for an individual. Assuming that the repository contained all the waste from the EEF Reference Reactor, using either conventional steel or a vanadium alloy as the first wall and blanket structural material, the maximum doses were estimated to be 2 x 10-i’ or 3 x lo-” Sv/y respectively. These maximum doses occur between lo5 and lo6 years after disposal, and are five orders of magnitude below the UK regulatory target; even with pessimistic. assumptions of unfavourable chemistry and a short ground-water return time the maximum dose rate is three orders of magnitude below the regulatory target. For comparison the same calculation was undertaken for the waste from a PWR, giving a maximum dose of 9 x lo-* sv/y. Although the above study did not specifically estimate the dose rates from gases, the special case of tritium was considered. The quantity of tritium in fusion waste, without additional processing, would be in excess of the present estimate of tritium from other sources in the UK National Inventory. On the basis of safety assessmentsof repositories it is not expected that tritium in fusion waste would lead to dose rates above the regulatory target, but are close enough that either

46

R. Hancox,

G.J. Buttenvorth

some additional processing may be necessaryor special precautions taken in the repository.

5. Conclusions

The waste which is expected to be produced by three generations of fusion devices, JET, NET and power reactors, has been considered. In the caseof the JET experiment, the materials used in construction are well known, and the total volume of waste at decommissioning is reasonably well known. The levels of activation of the waste have been estimated on the basis of the maximum credible number of neutrons which will be produced during operation, but the division of the waste into the categories of Intermediate, Low and Very Low Level waste will depend on the actual outcome of the experimental programme. Operational waste which cannot be disposed of as Low Level Waste during operation has been included in the decommissioning waste, but is a very small contribution. Preliminary studies of the decommissioning of JET indicate that the processesof dismantling the machine and packaging the waste can be undertaken with either the existing JET remote handling equipment or equipment similar to that used in other decommissioning projects. However, since JET will be the first fusion device which requires decommissioning and waste disposal, the successof these activities will be important for the continuing development of fusion as an environmentally attractive energy source and lessons will be learnt which should be applied in the design and operation of subsequent machines. In particular, the questions of the high levels of tritium in some components and of organic materials in others will need special attention. In the case of fusion reactors, the materials used in their construction are less well known and in particular will depend on the development of low activation materials for use in the first wall, blanket and shield. On the basis of assumptions about possible materials, the volumes and levels of activation can be calculated for a tokamak reactor which is sufficiently compact to be economically competitive. The volume of waste from the fusion reactor itself can be estimated with an uncertainty of a factor 3 and, although it will be greater than the equivalent volume from a PWR fission reactor, when included with the operational waste and waste from external components the volumes from the fusion system will be comparable to the volume from a PWR without fuel reprocessing. The volume from the fusion

/ The management

o//&on

waste

system will be less,however, than from a PWR with fuel reprocessing. From the viewpoint of public acceptance, a significant characteristic of the waste from fusion reactors is the health risk which it poses to future generations after disposal. Several estimates of the risks from fusion waste in shallow and deep repositories have been published, taking into account both intrusion and normal development. The results depend on the materials assumed and have wide uncertainties due to the lack of some appropriate data, but all show substantially lower risks than those from fission reactors, which are themselves very low. For materials buried in deep repositories, the risks to the public will be well below the levels set by current regulations. Some materials, especially large volumes of materials from the shield and external circuits or containing predominantly isotopes with short half-lives, could probably be disposed of safely and more economically in shallower repositories. This possibility, however, would depend on a better knowledge of the materials involved, improved safety cases and on changes in regulations in some countries. The disposal of nuclear waste has become a subject of considerable public and political concern. Fusion reactors do not include transuranic elements nor generate fission products, and the problems of waste disposal are correspondingly less than those for fission reactors. Nevertheless, fusion reactors do produce neutron activated structural and breeder materials and tritium. The future acceptance of fusion as a new power source will depend, in part, on minimizing these waste materials and clearly demonstrating that the long-term risks which they pose are socially acceptable.

References

[l] G.J. Butterworth, Low activation structural materialsfor fusion, Proceedingsof the 15th Symposiumon Fusion Technology,Utrecht, September1988. [2] M.J. Plewset al., The cost benefit of recyclinglow activation steelsfrom fusion reactors,CLM-R296 (August1989). [3] Preliminarydecommissioningplan for JET, Issue 2 (15 January1990). [4] C. Ponti, Activation calculationsfor NET-shieldingblanket, Ispra TechnicalNote No 1.88.73(1988). [5] P. Guttat, Fusionreactorwastes:Technicaland radiological aspectsfor the managementof wastesfrom NET and a commercialreactor, IAEA technical meeting on Fusion ReactorSafety,Culham, 3-7 November1987. [6] A. Hultgren and G. Olsson,NET wastedisposalin existing and planned repositories,IAEA technicalmeetingon Fusion ReactorSafety,Jackson,4-7 April 1989.

R. Hancox, G.J. Buttenvorth / The management ofjusion waste [7] R.S. Pease .et al., Environmental, safety-related and economic potential of fusion power (December 1989). [8] M.G. Sowerby and R.A. Forrest, A study of the environmental impact of fusion, AERE R 13708 (March 1990). [9] J.P. Holdren et al., Report of the Senior Committee on Environmental, Safety and Economic Aspects of Magnetic Fusion Energy, UCRL-53766 (September 1989).

41

[lo] K.R. Smith and G.J. Butterworth, IAEA technical meeting on Fusion Reactor Safety, Jackson, 4-7 April 1989. [ll] F. Aggeryd and V. Bergstrom, Doses to man due to leakage into the biosphere of fusion waste nuclides disposed of in a repository, NS-90/11 (January 1990).