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Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants
Fusion Engineering and Design 36 (1997) 49 – 56
Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants Karin Brode´n a,*, Maria Lindberg a, Simon Nisan 1,b, Paolo Rocco c, Massimo Zucchetti d, Neill Taylor e, Cleve Forty e a
Association EURATOM, Studs6ik RadWaste AB, S-611 82 Nyko¨ping, Sweden The NET Team, Boltzmannstraße 2, D-85 748, Garching bei Mu¨nchen, Germany c European Commission, Institute for Ad6anced Materials, Joint Research Centre, I-21 020 Ispra (VA), Italy d Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10 129 Torino, Italy e Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB, UK b
Abstract An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor (ITER) and also for future fusion power station designs currently under investigation within the European programme on the Safety and Environmental Assessment of fusion power, Long-term (SEAL). © 1997 Elsevier Science S.A.
1. Introduction The operation and decommissioning of fusion power plants will generate radioactive material due to neutron activation and tritium contamination from the deuterium/tritium fuel. Part of this material will also be chemically toxic. Management of waste will be an important issue for public opinion when deciding on the realisation and siting of the facilities. * Corresponding author. 1 Present address: DER/SIS, CEA/CEN Cadarache, F13108, Saint Paul-lez-Durance, France.
For several years fusion waste studies (for example quantification and qualification studies) and fusion waste related studies (for example material studies) have been performed within the European Fusion Technology Program. The present paper gives a survey of some important waste aspects that have been studied in Sweden, the UK, Italy and the Joint Research Centre of the European Commission. Results for both the International Thermonuclear Experimental Reactor (ITER) and future fusion power station designs investigated within the Safety and Environmental Assessment of fusion power, Long-term (SEAL) programme are presented.
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K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
More detailed recent results on ITER waste management strategies and final disposal are given in a poster presentation at this conference [1]. 2. Use of low-activation materials By appropriate selection of structural materials, it seems possible to reduce the active inventory, particularly that of the first wall, divertor and blanket. The use of specially-developed low-activation materials, which have a composition optimised to reduce neutron activation, is likely to be of benefit. It is important to note that this optimisation, reducing the active inventory on the time scale relevant to waste management, is not the same as optimisation to reduce activation and decay heat on the short time scale relevant to postulated accident scenarios. Furthermore, it is essential to appreciate the contribution that activated impurities in the composition may have on the long time scale. In order to illustrate these points, Fig. 1 shows some results for four candidate structural materials obtained with the inventory code FISPACT-4 [2] and the EAF4.1 cross section library [3]. The quantity plotted is the ingestion hazard potential, a biological hazard index important in waste management considerations because of the possible transport of radio-nuclides to the biosphere on geological time spans. Plots of other activation hazard indices show similar behaviour to this one.
Fig. 1. Ingestion dose hazard potential for four candidate structural materials.
The four materials compared are: V-4%Ti–3.3%Cr, a vanadium alloy with nominal composition [4]; LA12TaLC, a low-activation martensitic steel, considered in the SEAFP study [5]; F-82H, an alternative low-activation martensitic steel [6]; silicon carbide composite (SiC/SiC) [4]. All materials contain realistic impurity concentrations. The calculations represent the exposure of each material for 2.5 years in the blanket of a typical fusion power plant with a mean neutron wall loading of 2.0 MW m − 2, based on the design developed in the Safety and Environmental Assessment of Fusion Power (SEAFP) and the subsequent decay up to 105 years after shutdown. Fig. 1 shows that the ranking of these four materials in terms of this hazard index is markedly different at the long time scale compared with the short time scale. The better performers after about 50 years tend to be the poorer performers before that time and vice versa, although it is important to note that all four materials are better than stainless steel on all time scales, often substantially so. This conclusion is confirmed when considering other activation indices such as dose rate, decay power and specific activity, which also show the same trend. A further observation is the critical importance of impurities to the long time response of materials. Impurities present in parts per million concentrations may have a dominating influence on activation response at waste disposal time scales. This is illustrated by a comparison of the two steels, LA12TaLC and F-82H. Fig. 1 shows that within the nuclear data uncertainties, the results for both steels are coincident up to about 70 years. At times longer than this, the F-82H steel is consistently superior to LA12TaLC steel. Examination of the detailed results reveals that this is due to a reduced nitrogen content from 200 to 80 ppm. The important nuclide in this case is carbon14, produced through the reaction channel [14N](n,p)[14C]. In addition, other impurities become significant when different activation indices are considered. For example, reduction of the niobium impurity content from 50 ppm in LA12TaLC steel to 1 ppm in F-82H, results a
K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
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fifty-fold reduction in the gamma dose rate after 100 years in the F-82H steel compared with the LA12TaLC.
3. Detritiation From the waste management point of view, tritium has to be recovered to reduce outgassing during the interim storage and to comply with the regulatory limits for disposal. Of course, there are also powerful economic reasons for tritium recovery. Results from detritiation experiments in France indicate that the tritium content in tritium containing metals can be reduced to less than 500 GBq t − 1 [7]. These results were obtained in a small-scale, one-step industrial melting process for metal parts with initial tritium contents of up to about 30 000 GBq t − 1. The residual tritium content after detritiation will be further decreased during interim storage, due to decay. A specific activity of 500 GBq t − 1 will be reduced to about 30 GBq t − 1 after 50 years of interim storage. This figure compares favourably with the tritium limit for waste to the German repository Konrad. The tritium content in metal waste to Konrad must not exceed 190 GBq per package [8]. That means about six tonnes per package.
Fig. 2. The effect of intermediate storage on the destination of the waste from decommissioning the vacuum vessel of ITER in a hypothetical German scenario.
requiring waste to be sent to Gorleben is that the surface dose rate of the waste packages exceeds 2 mSv h − 1; another is that it is heat generating. A storage building is sufficient for interim storage of low-level waste before clearance. However, the highly activated waste has to be cooled in a dry or wet storage facility. An interim storage facility like the Swedish CLAB (central storage for spent fuel) could be used. Waste with high tritium content may require special precautions during storage for retaining the tritium inside the packages, e.g. tight welding, in combination with effective ventilation and air cleaning in the interim store.
4. Interim storage By interim storage the activity in the material is reduced before re-cycling, clearance, or final waste disposal. This can be illustrated by a hypothetical German scenario. Fig. 2 shows that increasing the intermediate storage time allows the waste from the vacuum vessel of ITER to be re-designated from the Gorleben to the Konrad repositories. Gorleben is a deep repository in a salt formation, Konrad is a repository in a former iron mine for non-heat generating waste delivered in packages that can be handled without shielding. It is clearly seen that if the storage time is prolonged a larger part of the waste can go to a less sophisticated repository. The main reason for
5. Recycling and clearance Assessments on recycling and clearance have been performed in the frame of SEAL [9] for the two power reactor designs, SEAFP Model 1 and 2 (see Table 1). Most of the detailed recycling and clearance assessments concerned SEAFP Model 2, the configuration adopting more conventional technological solutions. Additional results were produced for SEAFP Model 1, essentially for sake of comparison. Activation data were computed with the FISPACT-4 activation code [2]. The irradiation was
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K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
Table 1 Material in different components of SEAFP Model 1 and 2 Component
SEAFP Model 1
SEAFP Model 2
Blanket In-vessel components structure Divertor
Li2O V-5Ti Beryllium-armour V-5Ti heat sink Helium as coolant OPSTAB and water OPSTAB, water, lead and boron carbide AISI316LN structure Nb-Sn superconductor Copper conductor Helium as coolant Glass and epoxy insulator
Water cooled Pb-17Li LA12TaLC Beryllium-armour Copper heat sink Water as coolant OPSTAB and water OPSTAB, water, lead and boron carbide AISI316LN structure Nb-Sn superconductor Copper conductor Helium as coolant Glass and epoxy insulator
Shield Vessel Toroidal field coils
assumed to be done at the mean neutron wall loading of 2 MW m − 2. A fluence of about 10 MWa m − 2 for the first wall has been considered. Five and 25 years of continuous irradiation were assumed for the in-vessel components and the other zones, respectively. The activation data for the divertor were taken from previous SEAFP analyses. A 15-month irradiation was assumed for the Cu-AISI 316 divertor. All activation levels were computed after 50 years of storage, assumed to be the interim storage period at the plant site. Concerning recycling (reuse in the fusion industry), surface dose limits of 10 mSv h − 1 and 10 mSv h − 1 were assumed respectively for hands-on (HOR) and remote handling (RHR) recycling. These value are modified with respect to those adopted in SEAFP [5], where a surface dose rate limit of 25 mSv h − 1 was assumed for hand-on recycling and 2 mSv h − 1 were taken as limits for recycling by remote handling. The feasibility of clearance (declassification to non-active waste) was evaluated weighting the potential hazard of each radionuclide contributing to the radioactivity concentration of the examined material. This approach differs from that adopted in previous analyses for SEAFP [5] in SEAFP report where the global limit for clearance was 400 Bq kg − 1. This is the limit for Very Low Level Waste currently adopted in Britain and a very low value, compared with the natural radioactivity of many
substances, e.g. fertilisers and bricks which may exceed 5000 and 1000 Bq kg − 1, respectively. In the present analyses clearance levels related to each radio-nuclide were taken from an IAEA draft report [10]. In this document, clearance levels for many radionuclides are obtained from a categorisation of various safety analyses of radioactive waste repositories. These analyses assume 10 and 100 mSv year − 1 as dose limits for the most exposed individual in ‘likely’ and ‘unlikely’ accident scenarios happening in the waste repository. The clearance levels of radionuclides not included in [10] were evaluated with a fitting formula, taking the lowest value (Bq g − 1) from the three formulae: 1 ALIinh ALIing , , (1) Eg + 0.1Eb 1000 100,000 where Eg and Eb are the effective energies (MeV) of the gamma and beta emission, ALIinh and ALIing (Bq) are the most restrictive values of the Annual Limits of Intake by inhalation and ingestion. A ‘clearance index’ Ic was evaluated for each material, taking into account the contribution of all radionuclides contained. Ic was evaluated from the specific activity Ai and the clearance level Li of each one of the z radionuclides contained in the material: Z Ai (2) Ic = % i = 1 Li
!
"
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Table 2 Clearance indices and clearance feasibility for SEAFP Model 2 components Component
Cooling time (years)
Clearance index
Clearance feasibility
Inboard+outboard shield Outboard vessel Inboard vessel Outboard magnets Inboard magnets
50 10 50 50 50
1 0.9 1 0.02 1
NO YES NO YES YES
It is postulated that the material can be cleared if Ic 51. Typical clearance indices and the related clearance feasibility for SEAFP Model 2 components are shown in Table 2. Note that the outboard part of the vessel can be cleared after only 10 years of decay. Fig. 3 shows fractions of SEAFP Model 1 and 2 activated materials to waste disposal, remote handling recycling, hands-on recycling and nonactive waste. The activated material arising from SEAFP Model 2, is about 70 000 t. This amount includes all structures from the plasma chamber to the magnet zones and takes into account operation, maintenance and decommissioning. Forty-eight percent can be recycled, 39% can be cleared if the Ic are applied, 13% needs to be disposed of as active waste. An extension of the cooling time up to 100 years for the inner components could allow the recycling also of this fraction.
Fig. 3. Fractions of Model 1 and Model 2 materials to waste disposal, remote handling recycling, hands on recycling and non-active.
Appropriate detritiation procedures of the in vessel components before interim storage can reduce tritium inventories and tritium outgassing rates to such low levels as to not hinder recycling. Similarly, out-vessel material, subjected to occasional tritium contamination, could be cleared. The activated material arising from SEAFP Model 1, is about 60 000 t. However, from the radiological point of view, no activated waste needs to be disposed of, 70% of the material could be recycled (41% RHR, 29% HOR) and 30% could be cleared. The material which can be declassified is less than that in SEAFP Model 2, due to the better shielding characteristics of the SEAFP Model 2 Blanket.
6. Volume reduction The large waste components have to be segmented prior to further handling. For highly activated components segmenting techniques developed for fission reactor internals can be used. For example, the Japan Atomic Energy Research Institute (JAERI) has developed an underwater plasma arc technique for up to 130 mm thick stainless steel components [11]. Other techniques can be used for low level metallic waste. Oxy-acetylene, oxy-lance and plasma torches and also cold cutting with special purpose saws are examples of suitable techniques for this type of waste [12]. Melting can be used for volume reduction of metal waste and also as a treatment method for metals that can be recycled or declassified. By the melting procedure it is possible to get a homogenous distribution of the activity in the metal and to take out representative samples for analysis.
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used also for the ITER waste. For the present calculation of ITER waste, for Konrad repository in the German scenario, the container which gives the lowest volume is the SBOX-II. This package is a rectangular steel container with a volume of 4.6 m3. The packing efficiency is generally assumed to be 30% of the available inner volume for all materials. Fig. 5 shows the repository volume required for packages with permanent steel components of ITER. The blankets are not included. The mixed steel consists of the upper and lower crowns, the gravity support, the pre-load structure and the outer PF-coil support. The central solenoid can be declassificated to non-active waste and is therefore not included.
8. Waste disposal
Fig. 4. The melting facility for low-level carbon steel scrap and low-level stainless scrap at Studsvik.
Fig. 4 shows the melting facility at Studsvik RadWaste. Incineration can be used for volume reduction of burnable dry active waste.
Potential doses to man from ITER waste in German type repositories have been calculated based on available results from safety assessments for the German repositories Gorleben and Konrad. Gorleben is a deep geological repository in a salt formation. Two methods for disposal are expected to be used in the repository: tunnel disposal for non-heat generating waste at 870 m level and possibly also at 900 m; bore-hole disposal at 300–600 m from the 870 m level for heat generating waste.
7. Packaging When putting waste into packagings there are some factors that set the destination and other factors that limit the packages and control the amount of waste. This can be exemplified by a hypothetical German scenario with the deep repositories Konrad and Gorleben. The maximum weight for Konrad is 20 t and for Gorleben 40 t and the maximum surface dose rate for Konrad is 2 mSv h − 1 and for Gorleben 500 mSv h − 1. When calculating the volume of packaged waste the standard containers used for fission waste in the repository today have been assumed to be
Fig. 5. The repository volumes of the permanent steel components of ITER for a hypothetical German scenario.
K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
A remote handling system will be used for the bore-hole disposal. Konrad is a deep geological repository (800– 1300 m) in a former iron mine. It is a repository for non-heat generating waste delivered in packages that can be handled without shielding. The safety assessment for Gorleben includes calculated dose rate values for Ni-59, Tc-99 and Nb-94 [13]. It has been assumed that ground-water could intrude to the repository area of Gorleben in the post-operational phase and that there could occur migration of radionuclide with ground-water to surface-water. The results were used to estimate the dose equivalent rate for Ni-59, Tc-99 and Nb-94 in ITER waste. If, for example, the shielding blanket of ITER is placed in Gorleben the maximum dose equivalent rate from Ni-59, Tc-99 and Nb-94 is estimated to be 1.3× 10 − 9, 3.9 × 10 − 10 and 3.8 × 10 − 11 Sv year − 1, respectively. Dose rate calculation results are also available for Konrad [14]. The transport times from the repository to the biosphere were extremely long in these calculations. The major parts of the radioactive nuclides will therefore decay before they could reach man. Tc-99 gave a maximum dose equivalent rate after about three million years. If the waste from all permanent steel components inside the biological shield, (i.e. excluding the blanket and free components), is placed in Konrad and if the migration rate of Ni-59 and Nb-94 is estimated to be in the same order of magnitude as Tc-99, only 4.7 × 106 Bq Tc-99, 4 Bq Ni-59 and nothing of Nb-94 will remain after three million years and the dose rate from the nuclides will thus be insignificant.
55
Fig. 6. A comparison between the radiological and the chemical ALI for the shielding blanket of ITER.
pared for the shielding blanket of the ITER machine, assuming beryllium is used as the armour material throughout. The blanket consists of about 7000 tonnes of SS316LN steel and about 10 t of beryllium. The radiological ALI was in this case calculated as the sum of the number of ALIs of the nuclides, Cr-51, Fe-55, Ni-59, Ni-63 and Co-60 in the shielding blanket while the chemical ALI was derived from the occupational exposure limit values given in units of mg m − 3 multiplied by the volume of air inhaled per year by one person (which is 2400 m3). As seen in Fig. 6 the radiological ALI decreases and 250 years after shut-down is lower than the chemical ALI for Be. The chemical ALI on the other hand does not change with time and in the long run it will dominate the hazard of the ITER waste. 10. Conclusions
9. Chemically toxic waste Depending on the materials used in the different parts of a fusion reactor the waste might not only be activated it may also be toxic. One of the most toxic materials in some designs is beryllium which may be used as armour of the plasma-facing surfaces and as a neutron multiplier in blankets. In Fig. 6 the radiological ALI (Annual Limits on Intake) and the chemical ALI are com-
Use of low activation structural materials greatly reduces the content of long-lived and hazardous nuclides in the activated material. A vanadium alloy, two different low-activation martensitic steels and silicon carbide have been studied. The hazard potential from all these materials after irradiation will be substantially lower than from stainless steel. One of low-activation martensitic steels, F-82H, will give the lowest hazard potential after about 100 years of decay.
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Tritium is a valuable fusion fuel and should be extracted from the fusion waste. Also for regulatory reasons it is important to reduce the tritium content in the waste. By detritiation followed by 50 years of interim storage it is possible to reduce the tritium concentration below the limit set for the German repository Konrad, for example. Clearance can be used to reduce the quantity of material finally classified as very low-level waste. In the future it may also be possible to reduce the quantity of waste with higher activity contents by recycling of materials with dose rates up to 10 mSv h − 1. This means that almost no material from future power stations may need to be disposed of in a waste repository. However, part of the material from ITER will need to be disposed of in a repository: a less sophisticated one such as the German repository Konrad seems to be adequate.
References [1] S. Nisan, K. Brode´n and M. Lindberg, ITER Waste Management Strategies and Final Disposal, in: Fusion Technology 1996, Proceedings of the 19th Symposium on Fusion Technology, Lisbon, 16–20 September, 1996, North-Holland, Elsevier, Amsterdam, 1997. [2] R.A. Forrest and J.-Ch. Sublet, FISPACT4 User Manual, UKAEA report UKAEA FUS 287, 1995.
.
[3] J. Kopecky and D. Nierop, The European Activation File EAF-4, ECN Petten report ECN-C-95-072, 1995. [4] W. Dietz, private communication with I. Cook/N. Taylor, November 1995. [5] J. Raeder et al. ‘Safety and Environmental Assessment of Fusion Power’, European Commission report EURFUBRU XII-217/95, 1995. [6] N. Yamanouchi, et al., J. Nucl. Mater. 191-194 (1992) 822. [7] L. Boisset and C. Lattaud, Metallic waste detritiation. Performance of a melting process. CEA Note Technique 95/039, 04/08/95. [8] P. Brennecke and E. Warnecke, Anforderungen an endzulagernde radioaktive Abfa¨lle (Vorla¨ufige Endlagerungsbedingungen, Stand April 1990 in der Fassung Juli 1991). -Schachtanlage Konrad. Bundesamt fu¨r Strahlenschutz ET-3/90-REV-1, 1991. [9] P. Rocco and M. Zucchetti, Recycling and Clearance of Fusion Waste, final report of SEAL 10.1, JRC Ispra, SEALAG, August 1996. EUR 16453 EN. [10] Clearance Levels for Radionuclides in Solid Materials: Application of Exemption Principles, IAEA Draft Safety Guide, Safety Series No. 111 G 1 – 5, Vienna, 1994. [11] S. Yanagihara, Y. Seiki, H. Nakamura, Dismantling Techniques for Reactor Steel Stuctures, Nucl. Techn. 86 (1989) 148 – 158. [12] O. Andersson, Minimising Low Level Waste by Volume Reduction and Recycling. Internationale Zeitschrift fu¨r Kernenergie atw 40 Jg (1995) Heft 7 – Juli, 461 – 465. [13] Entwicklung eines sicherheitsanalytischen Instrumentariums fu¨r das geologische Endlager fu¨r radioaktive Abfa¨lle in einem Salzstock, Zusammenfassender Abschlußbericht, Kapitel 4, Projekt Sicherheitsstudien Entsorgung, Berlin, 1985. [14] H. Illi, Safety for the planned Konrad repository, Chemietechnik SI (1987) No. 2.
Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants Karin Brode´n a,*, Maria Lindberg a, Simon Nisan 1,b, Paolo Rocco c, Massimo Zucchetti d, Neill Taylor e, Cleve Forty e a
Association EURATOM, Studs6ik RadWaste AB, S-611 82 Nyko¨ping, Sweden The NET Team, Boltzmannstraße 2, D-85 748, Garching bei Mu¨nchen, Germany c European Commission, Institute for Ad6anced Materials, Joint Research Centre, I-21 020 Ispra (VA), Italy d Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10 129 Torino, Italy e Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB, UK b
Abstract An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor (ITER) and also for future fusion power station designs currently under investigation within the European programme on the Safety and Environmental Assessment of fusion power, Long-term (SEAL). © 1997 Elsevier Science S.A.
1. Introduction The operation and decommissioning of fusion power plants will generate radioactive material due to neutron activation and tritium contamination from the deuterium/tritium fuel. Part of this material will also be chemically toxic. Management of waste will be an important issue for public opinion when deciding on the realisation and siting of the facilities. * Corresponding author. 1 Present address: DER/SIS, CEA/CEN Cadarache, F13108, Saint Paul-lez-Durance, France.
For several years fusion waste studies (for example quantification and qualification studies) and fusion waste related studies (for example material studies) have been performed within the European Fusion Technology Program. The present paper gives a survey of some important waste aspects that have been studied in Sweden, the UK, Italy and the Joint Research Centre of the European Commission. Results for both the International Thermonuclear Experimental Reactor (ITER) and future fusion power station designs investigated within the Safety and Environmental Assessment of fusion power, Long-term (SEAL) programme are presented.
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K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
More detailed recent results on ITER waste management strategies and final disposal are given in a poster presentation at this conference [1]. 2. Use of low-activation materials By appropriate selection of structural materials, it seems possible to reduce the active inventory, particularly that of the first wall, divertor and blanket. The use of specially-developed low-activation materials, which have a composition optimised to reduce neutron activation, is likely to be of benefit. It is important to note that this optimisation, reducing the active inventory on the time scale relevant to waste management, is not the same as optimisation to reduce activation and decay heat on the short time scale relevant to postulated accident scenarios. Furthermore, it is essential to appreciate the contribution that activated impurities in the composition may have on the long time scale. In order to illustrate these points, Fig. 1 shows some results for four candidate structural materials obtained with the inventory code FISPACT-4 [2] and the EAF4.1 cross section library [3]. The quantity plotted is the ingestion hazard potential, a biological hazard index important in waste management considerations because of the possible transport of radio-nuclides to the biosphere on geological time spans. Plots of other activation hazard indices show similar behaviour to this one.
Fig. 1. Ingestion dose hazard potential for four candidate structural materials.
The four materials compared are: V-4%Ti–3.3%Cr, a vanadium alloy with nominal composition [4]; LA12TaLC, a low-activation martensitic steel, considered in the SEAFP study [5]; F-82H, an alternative low-activation martensitic steel [6]; silicon carbide composite (SiC/SiC) [4]. All materials contain realistic impurity concentrations. The calculations represent the exposure of each material for 2.5 years in the blanket of a typical fusion power plant with a mean neutron wall loading of 2.0 MW m − 2, based on the design developed in the Safety and Environmental Assessment of Fusion Power (SEAFP) and the subsequent decay up to 105 years after shutdown. Fig. 1 shows that the ranking of these four materials in terms of this hazard index is markedly different at the long time scale compared with the short time scale. The better performers after about 50 years tend to be the poorer performers before that time and vice versa, although it is important to note that all four materials are better than stainless steel on all time scales, often substantially so. This conclusion is confirmed when considering other activation indices such as dose rate, decay power and specific activity, which also show the same trend. A further observation is the critical importance of impurities to the long time response of materials. Impurities present in parts per million concentrations may have a dominating influence on activation response at waste disposal time scales. This is illustrated by a comparison of the two steels, LA12TaLC and F-82H. Fig. 1 shows that within the nuclear data uncertainties, the results for both steels are coincident up to about 70 years. At times longer than this, the F-82H steel is consistently superior to LA12TaLC steel. Examination of the detailed results reveals that this is due to a reduced nitrogen content from 200 to 80 ppm. The important nuclide in this case is carbon14, produced through the reaction channel [14N](n,p)[14C]. In addition, other impurities become significant when different activation indices are considered. For example, reduction of the niobium impurity content from 50 ppm in LA12TaLC steel to 1 ppm in F-82H, results a
K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
51
fifty-fold reduction in the gamma dose rate after 100 years in the F-82H steel compared with the LA12TaLC.
3. Detritiation From the waste management point of view, tritium has to be recovered to reduce outgassing during the interim storage and to comply with the regulatory limits for disposal. Of course, there are also powerful economic reasons for tritium recovery. Results from detritiation experiments in France indicate that the tritium content in tritium containing metals can be reduced to less than 500 GBq t − 1 [7]. These results were obtained in a small-scale, one-step industrial melting process for metal parts with initial tritium contents of up to about 30 000 GBq t − 1. The residual tritium content after detritiation will be further decreased during interim storage, due to decay. A specific activity of 500 GBq t − 1 will be reduced to about 30 GBq t − 1 after 50 years of interim storage. This figure compares favourably with the tritium limit for waste to the German repository Konrad. The tritium content in metal waste to Konrad must not exceed 190 GBq per package [8]. That means about six tonnes per package.
Fig. 2. The effect of intermediate storage on the destination of the waste from decommissioning the vacuum vessel of ITER in a hypothetical German scenario.
requiring waste to be sent to Gorleben is that the surface dose rate of the waste packages exceeds 2 mSv h − 1; another is that it is heat generating. A storage building is sufficient for interim storage of low-level waste before clearance. However, the highly activated waste has to be cooled in a dry or wet storage facility. An interim storage facility like the Swedish CLAB (central storage for spent fuel) could be used. Waste with high tritium content may require special precautions during storage for retaining the tritium inside the packages, e.g. tight welding, in combination with effective ventilation and air cleaning in the interim store.
4. Interim storage By interim storage the activity in the material is reduced before re-cycling, clearance, or final waste disposal. This can be illustrated by a hypothetical German scenario. Fig. 2 shows that increasing the intermediate storage time allows the waste from the vacuum vessel of ITER to be re-designated from the Gorleben to the Konrad repositories. Gorleben is a deep repository in a salt formation, Konrad is a repository in a former iron mine for non-heat generating waste delivered in packages that can be handled without shielding. It is clearly seen that if the storage time is prolonged a larger part of the waste can go to a less sophisticated repository. The main reason for
5. Recycling and clearance Assessments on recycling and clearance have been performed in the frame of SEAL [9] for the two power reactor designs, SEAFP Model 1 and 2 (see Table 1). Most of the detailed recycling and clearance assessments concerned SEAFP Model 2, the configuration adopting more conventional technological solutions. Additional results were produced for SEAFP Model 1, essentially for sake of comparison. Activation data were computed with the FISPACT-4 activation code [2]. The irradiation was
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K. Brode´n et al. / Fusion Engineering and Design 36 (1997) 49–56
Table 1 Material in different components of SEAFP Model 1 and 2 Component
SEAFP Model 1
SEAFP Model 2
Blanket In-vessel components structure Divertor
Li2O V-5Ti Beryllium-armour V-5Ti heat sink Helium as coolant OPSTAB and water OPSTAB, water, lead and boron carbide AISI316LN structure Nb-Sn superconductor Copper conductor Helium as coolant Glass and epoxy insulator
Water cooled Pb-17Li LA12TaLC Beryllium-armour Copper heat sink Water as coolant OPSTAB and water OPSTAB, water, lead and boron carbide AISI316LN structure Nb-Sn superconductor Copper conductor Helium as coolant Glass and epoxy insulator
Shield Vessel Toroidal field coils
assumed to be done at the mean neutron wall loading of 2 MW m − 2. A fluence of about 10 MWa m − 2 for the first wall has been considered. Five and 25 years of continuous irradiation were assumed for the in-vessel components and the other zones, respectively. The activation data for the divertor were taken from previous SEAFP analyses. A 15-month irradiation was assumed for the Cu-AISI 316 divertor. All activation levels were computed after 50 years of storage, assumed to be the interim storage period at the plant site. Concerning recycling (reuse in the fusion industry), surface dose limits of 10 mSv h − 1 and 10 mSv h − 1 were assumed respectively for hands-on (HOR) and remote handling (RHR) recycling. These value are modified with respect to those adopted in SEAFP [5], where a surface dose rate limit of 25 mSv h − 1 was assumed for hand-on recycling and 2 mSv h − 1 were taken as limits for recycling by remote handling. The feasibility of clearance (declassification to non-active waste) was evaluated weighting the potential hazard of each radionuclide contributing to the radioactivity concentration of the examined material. This approach differs from that adopted in previous analyses for SEAFP [5] in SEAFP report where the global limit for clearance was 400 Bq kg − 1. This is the limit for Very Low Level Waste currently adopted in Britain and a very low value, compared with the natural radioactivity of many
substances, e.g. fertilisers and bricks which may exceed 5000 and 1000 Bq kg − 1, respectively. In the present analyses clearance levels related to each radio-nuclide were taken from an IAEA draft report [10]. In this document, clearance levels for many radionuclides are obtained from a categorisation of various safety analyses of radioactive waste repositories. These analyses assume 10 and 100 mSv year − 1 as dose limits for the most exposed individual in ‘likely’ and ‘unlikely’ accident scenarios happening in the waste repository. The clearance levels of radionuclides not included in [10] were evaluated with a fitting formula, taking the lowest value (Bq g − 1) from the three formulae: 1 ALIinh ALIing , , (1) Eg + 0.1Eb 1000 100,000 where Eg and Eb are the effective energies (MeV) of the gamma and beta emission, ALIinh and ALIing (Bq) are the most restrictive values of the Annual Limits of Intake by inhalation and ingestion. A ‘clearance index’ Ic was evaluated for each material, taking into account the contribution of all radionuclides contained. Ic was evaluated from the specific activity Ai and the clearance level Li of each one of the z radionuclides contained in the material: Z Ai (2) Ic = % i = 1 Li
!
"
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Table 2 Clearance indices and clearance feasibility for SEAFP Model 2 components Component
Cooling time (years)
Clearance index
Clearance feasibility
Inboard+outboard shield Outboard vessel Inboard vessel Outboard magnets Inboard magnets
50 10 50 50 50
1 0.9 1 0.02 1
NO YES NO YES YES
It is postulated that the material can be cleared if Ic 51. Typical clearance indices and the related clearance feasibility for SEAFP Model 2 components are shown in Table 2. Note that the outboard part of the vessel can be cleared after only 10 years of decay. Fig. 3 shows fractions of SEAFP Model 1 and 2 activated materials to waste disposal, remote handling recycling, hands-on recycling and nonactive waste. The activated material arising from SEAFP Model 2, is about 70 000 t. This amount includes all structures from the plasma chamber to the magnet zones and takes into account operation, maintenance and decommissioning. Forty-eight percent can be recycled, 39% can be cleared if the Ic are applied, 13% needs to be disposed of as active waste. An extension of the cooling time up to 100 years for the inner components could allow the recycling also of this fraction.
Fig. 3. Fractions of Model 1 and Model 2 materials to waste disposal, remote handling recycling, hands on recycling and non-active.
Appropriate detritiation procedures of the in vessel components before interim storage can reduce tritium inventories and tritium outgassing rates to such low levels as to not hinder recycling. Similarly, out-vessel material, subjected to occasional tritium contamination, could be cleared. The activated material arising from SEAFP Model 1, is about 60 000 t. However, from the radiological point of view, no activated waste needs to be disposed of, 70% of the material could be recycled (41% RHR, 29% HOR) and 30% could be cleared. The material which can be declassified is less than that in SEAFP Model 2, due to the better shielding characteristics of the SEAFP Model 2 Blanket.
6. Volume reduction The large waste components have to be segmented prior to further handling. For highly activated components segmenting techniques developed for fission reactor internals can be used. For example, the Japan Atomic Energy Research Institute (JAERI) has developed an underwater plasma arc technique for up to 130 mm thick stainless steel components [11]. Other techniques can be used for low level metallic waste. Oxy-acetylene, oxy-lance and plasma torches and also cold cutting with special purpose saws are examples of suitable techniques for this type of waste [12]. Melting can be used for volume reduction of metal waste and also as a treatment method for metals that can be recycled or declassified. By the melting procedure it is possible to get a homogenous distribution of the activity in the metal and to take out representative samples for analysis.
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used also for the ITER waste. For the present calculation of ITER waste, for Konrad repository in the German scenario, the container which gives the lowest volume is the SBOX-II. This package is a rectangular steel container with a volume of 4.6 m3. The packing efficiency is generally assumed to be 30% of the available inner volume for all materials. Fig. 5 shows the repository volume required for packages with permanent steel components of ITER. The blankets are not included. The mixed steel consists of the upper and lower crowns, the gravity support, the pre-load structure and the outer PF-coil support. The central solenoid can be declassificated to non-active waste and is therefore not included.
8. Waste disposal
Fig. 4. The melting facility for low-level carbon steel scrap and low-level stainless scrap at Studsvik.
Fig. 4 shows the melting facility at Studsvik RadWaste. Incineration can be used for volume reduction of burnable dry active waste.
Potential doses to man from ITER waste in German type repositories have been calculated based on available results from safety assessments for the German repositories Gorleben and Konrad. Gorleben is a deep geological repository in a salt formation. Two methods for disposal are expected to be used in the repository: tunnel disposal for non-heat generating waste at 870 m level and possibly also at 900 m; bore-hole disposal at 300–600 m from the 870 m level for heat generating waste.
7. Packaging When putting waste into packagings there are some factors that set the destination and other factors that limit the packages and control the amount of waste. This can be exemplified by a hypothetical German scenario with the deep repositories Konrad and Gorleben. The maximum weight for Konrad is 20 t and for Gorleben 40 t and the maximum surface dose rate for Konrad is 2 mSv h − 1 and for Gorleben 500 mSv h − 1. When calculating the volume of packaged waste the standard containers used for fission waste in the repository today have been assumed to be
Fig. 5. The repository volumes of the permanent steel components of ITER for a hypothetical German scenario.
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A remote handling system will be used for the bore-hole disposal. Konrad is a deep geological repository (800– 1300 m) in a former iron mine. It is a repository for non-heat generating waste delivered in packages that can be handled without shielding. The safety assessment for Gorleben includes calculated dose rate values for Ni-59, Tc-99 and Nb-94 [13]. It has been assumed that ground-water could intrude to the repository area of Gorleben in the post-operational phase and that there could occur migration of radionuclide with ground-water to surface-water. The results were used to estimate the dose equivalent rate for Ni-59, Tc-99 and Nb-94 in ITER waste. If, for example, the shielding blanket of ITER is placed in Gorleben the maximum dose equivalent rate from Ni-59, Tc-99 and Nb-94 is estimated to be 1.3× 10 − 9, 3.9 × 10 − 10 and 3.8 × 10 − 11 Sv year − 1, respectively. Dose rate calculation results are also available for Konrad [14]. The transport times from the repository to the biosphere were extremely long in these calculations. The major parts of the radioactive nuclides will therefore decay before they could reach man. Tc-99 gave a maximum dose equivalent rate after about three million years. If the waste from all permanent steel components inside the biological shield, (i.e. excluding the blanket and free components), is placed in Konrad and if the migration rate of Ni-59 and Nb-94 is estimated to be in the same order of magnitude as Tc-99, only 4.7 × 106 Bq Tc-99, 4 Bq Ni-59 and nothing of Nb-94 will remain after three million years and the dose rate from the nuclides will thus be insignificant.
55
Fig. 6. A comparison between the radiological and the chemical ALI for the shielding blanket of ITER.
pared for the shielding blanket of the ITER machine, assuming beryllium is used as the armour material throughout. The blanket consists of about 7000 tonnes of SS316LN steel and about 10 t of beryllium. The radiological ALI was in this case calculated as the sum of the number of ALIs of the nuclides, Cr-51, Fe-55, Ni-59, Ni-63 and Co-60 in the shielding blanket while the chemical ALI was derived from the occupational exposure limit values given in units of mg m − 3 multiplied by the volume of air inhaled per year by one person (which is 2400 m3). As seen in Fig. 6 the radiological ALI decreases and 250 years after shut-down is lower than the chemical ALI for Be. The chemical ALI on the other hand does not change with time and in the long run it will dominate the hazard of the ITER waste. 10. Conclusions
9. Chemically toxic waste Depending on the materials used in the different parts of a fusion reactor the waste might not only be activated it may also be toxic. One of the most toxic materials in some designs is beryllium which may be used as armour of the plasma-facing surfaces and as a neutron multiplier in blankets. In Fig. 6 the radiological ALI (Annual Limits on Intake) and the chemical ALI are com-
Use of low activation structural materials greatly reduces the content of long-lived and hazardous nuclides in the activated material. A vanadium alloy, two different low-activation martensitic steels and silicon carbide have been studied. The hazard potential from all these materials after irradiation will be substantially lower than from stainless steel. One of low-activation martensitic steels, F-82H, will give the lowest hazard potential after about 100 years of decay.
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Tritium is a valuable fusion fuel and should be extracted from the fusion waste. Also for regulatory reasons it is important to reduce the tritium content in the waste. By detritiation followed by 50 years of interim storage it is possible to reduce the tritium concentration below the limit set for the German repository Konrad, for example. Clearance can be used to reduce the quantity of material finally classified as very low-level waste. In the future it may also be possible to reduce the quantity of waste with higher activity contents by recycling of materials with dose rates up to 10 mSv h − 1. This means that almost no material from future power stations may need to be disposed of in a waste repository. However, part of the material from ITER will need to be disposed of in a repository: a less sophisticated one such as the German repository Konrad seems to be adequate.
References [1] S. Nisan, K. Brode´n and M. Lindberg, ITER Waste Management Strategies and Final Disposal, in: Fusion Technology 1996, Proceedings of the 19th Symposium on Fusion Technology, Lisbon, 16–20 September, 1996, North-Holland, Elsevier, Amsterdam, 1997. [2] R.A. Forrest and J.-Ch. Sublet, FISPACT4 User Manual, UKAEA report UKAEA FUS 287, 1995.
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[3] J. Kopecky and D. Nierop, The European Activation File EAF-4, ECN Petten report ECN-C-95-072, 1995. [4] W. Dietz, private communication with I. Cook/N. Taylor, November 1995. [5] J. Raeder et al. ‘Safety and Environmental Assessment of Fusion Power’, European Commission report EURFUBRU XII-217/95, 1995. [6] N. Yamanouchi, et al., J. Nucl. Mater. 191-194 (1992) 822. [7] L. Boisset and C. Lattaud, Metallic waste detritiation. Performance of a melting process. CEA Note Technique 95/039, 04/08/95. [8] P. Brennecke and E. Warnecke, Anforderungen an endzulagernde radioaktive Abfa¨lle (Vorla¨ufige Endlagerungsbedingungen, Stand April 1990 in der Fassung Juli 1991). -Schachtanlage Konrad. Bundesamt fu¨r Strahlenschutz ET-3/90-REV-1, 1991. [9] P. Rocco and M. Zucchetti, Recycling and Clearance of Fusion Waste, final report of SEAL 10.1, JRC Ispra, SEALAG, August 1996. EUR 16453 EN. [10] Clearance Levels for Radionuclides in Solid Materials: Application of Exemption Principles, IAEA Draft Safety Guide, Safety Series No. 111 G 1 – 5, Vienna, 1994. [11] S. Yanagihara, Y. Seiki, H. Nakamura, Dismantling Techniques for Reactor Steel Stuctures, Nucl. Techn. 86 (1989) 148 – 158. [12] O. Andersson, Minimising Low Level Waste by Volume Reduction and Recycling. Internationale Zeitschrift fu¨r Kernenergie atw 40 Jg (1995) Heft 7 – Juli, 461 – 465. [13] Entwicklung eines sicherheitsanalytischen Instrumentariums fu¨r das geologische Endlager fu¨r radioaktive Abfa¨lle in einem Salzstock, Zusammenfassender Abschlußbericht, Kapitel 4, Projekt Sicherheitsstudien Entsorgung, Berlin, 1985. [14] H. Illi, Safety for the planned Konrad repository, Chemietechnik SI (1987) No. 2.