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Waste management plans for ITER
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Waste management plans for ITER S. Rosanvallon ∗ , D. Torcy, J.K. Chon, A. Dammann ITER Organization, CS 90 046, 13067 St Paul lez Durance Cedex, France
a r t i c l e
i n f o
Article history: Received 30 July 2015 Received in revised form 30 November 2015 Accepted 2 December 2015 Available online xxx Keywords: ITER Nuclear Waste Characterization Decommissioning
a b s t r a c t ITER will produce radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning. The waste concerns components that are activated by neutrons of energies up to 14 MeV, and are contaminated by activated corrosion products, activated dust and tritium. Even if the nuclear waste production will start only with the deuterium-deuterium phase, provisions have to be taken now with the design of the ITER facilities that will be used for the treatment and interim storage of the waste, in order to demonstrate that all the ITER waste will be safely manageable with the existing outlets. This demonstration also needs to be provided to the French regulator. © 2015 Elsevier B.V. All rights reserved.
1. Introduction Neutrons will be produced by deuterium–deuterium and deuterium–tritium plasmas in the ITER plasma chamber and will lead to the activation of the structures. In the same time, these components will be contaminated by activated corrosion products, activated dust and tritium. During maintenance operations, some in-vessel components are anticipated to be removed from the vacuum vessel moreover supporting systems such as tritium fuel cycle systems or heating systems are planned to be removed. These components, once discarded, as well as those dismantled during decommissioning, will be treated as radioactive waste. Provisions have to be taken at the design stage of the ITER facilities that will be used for the treatment and interim storage of the waste, in order to demonstrate that all the ITER waste will be safely manageable with the existing disposals. This demonstration also needs to be provided to the French regulator. The aim of this paper is to describe the waste origin, the waste inventories, the optimization process put in place to reduce the waste radiotoxicity and volumes, and the overall strategy from component removal to final disposal that can be foreseen at this stage of the project. The main challenges of the waste treatments are also highlighted. It has to be noted that radioactive waste management follows in France the framework of a law on the sustainable management of
∗ Corresponding author. Tel.: +33 442176786. E-mail address: [email protected] (S. Rosanvallon).
radioactive materials and waste established in June 2006 [1]. Waste management is a Protection Important Activity as per 2012 Order [2].
2. Waste characteristics 2.1. Radioactive content After a non-radioactive phase using hydrogen plasmas, ITER, which is the Nuclear Facility INB-174, will operate with deuterium plasmas (D–D) and deuterium–tritium plasmas (D–T). The activation of the structures will start with the 2.5 MeV neutrons resulting from the D–D reactions and will continue with the 14 MeV neutrons arising from the D–T reaction. In addition to activation, components may be contaminated by tritium, activated and tritiated dust and activated corrosion products (ACPs). In fact, as soon as tritium is used as fuel for the fusion reaction, in-vessel components will be contaminated by tritium adsorption and permeation. Fuelling circuit components (fuelling system and tritium plant) will not be activated but will be tritiated due to tritium permeation. Moreover, the plasma interacts with the plasma facing components during steady-state operation and transients (mainly disruptions) leading to the creation of dust that will have the same radiological characteristics as the first few microns of the plasma facing components and will be thus both activated and tritiated. Furthermore, water circulating in the heat transfer systems will corrode the pipes, leading to activated corrosion products in the fluids and pipe circuits.
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Please cite this article in press as: S. Rosanvallon, et al., Waste management plans for ITER, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.12.002
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2.2. Radioactive waste As mentioned in [3], there is no exemption and/or clearance in France. The regulations impose that the nuclear operator defines a waste zoning of the facilities. It consists in defining zones producing radioactive waste and those producing non-radioactive waste. These zones are defined based on an analysis of the normal operations foreseen room by room and the potential exposure to activation and/or contamination. All the waste arising from a zone producing radioactive waste will have to be managed as radioactive waste as a function of the Andra (French national radioactive waste management agency) classification for radioactive waste. The operation and decommissioning of ITER will lead to gaseous effluents, liquids effluents and solid waste. The radioactive gaseous effluents are released to the environment via the ventilation and/or detritiation systems and through a monitored exhaust of the Tokamak Complex. They are not treated in this paper. The radioactive liquids effluents consists in effluents with radioactivity levels exceeding the limit values for discharge into the industrial sewage system and that need to be further processed to be considered as waste. The solid radioactive waste arises from day-to-day operation, maintenance and as a result of component replacement from radioactive waste zones. After treatment and solidification, liquid effluents lead also to solid waste. The solid waste classification is managed by the Andra. ITER radioactive waste can be classified into three categories based on its activity and radio-toxicity (nota: French abbreviations are used since they correspond to specific acceptation criteria): • TFA waste (French abbreviation for “Très Faible Activité” meaning very low-level), • FMA-VC waste (French abbreviation for “Faible et Moyenne Activité à Vie Courte” meaning low and intermediate-level short-lived) or type A waste, • MA-VL waste (French abbreviation for “Moyenne Activité à Vie Longue” meaning intermediate-level long-lived) or type-B waste. No HAVL waste (French abbreviation for “Haute Activité à Vie Longue” meaning high-level long-lived waste) is generated by the ITER facility. In addition to the ANDRA categories, because of the use of tritium as fuel in ITER, purely tritiated waste arising from the tritium plant for example, is also considered. This consideration is to enable a specific routing for this type of waste to avoid crosscontamination with waste that is activated and tritiated during interim storage. After radioactive decay, it is anticipated that this waste will be sent to the FMA-VC storage centre. The criteria for waste classification are given in [4]. 2.3. Inventories During the licensing process, the waste inventories are assessed to demonstrate that all the waste produced by the facility are manageable from the production (on-site treatments and interim storage) to the off-site treatments, if any, and final disposal. A first assessment has been prepared for the Preliminary Safety Report (RPrS) that has been submitted to the French Regulator in 2011. This assessment was established considering an inventory based on the design and the maintenance plans as they were standing and the classification was based on 1-D activation calculations. Since then, 3-D activation calculations have been carried out to reflect the latest updates on the materials used, the design of the machine and the planned maintenance operations. For the waste classification, tritium due to sources of trapping such as
implantation, recombination or permeation has been added to the tritium due to the activation. The classification of the main components at the end of the ITER lifetime is presented in Table 1. In some cases, the classification is given after a radioactive decay time after plasma shutdown. It concerns only components that are to be discarded during decommissioning. With these new calculations, the inner wall of the vacuum vessel is slightly above the acceptance criteria for FMA-VC waste. Nevertheless, when the vacuum vessel is considered globally (inner wall, shielding and outer wall all together), the classification remains FMA-VC as presented in Table 1. This considers also the presence of tritium in the waste. These results need to be further assessed as they could impact the decommissioning. Decommissioning scenarios were also proposed in the RPrS. They are being updated and based on the scenario that will be considered, it may be needed to reconsider this classification. A review of the ITER operational waste has also been performed in 2014. The operational waste consists in: • Process and housekeeping waste (gloves, papers, ion exchange resins, filters, oils, etc.) as in all nuclear facility, • Component replacement linked to the experimental program on plasma facing materials and to the maintenance of supporting systems such as tritium fuel cycle systems or heating systems, • Liquid effluents from the maintenance of the cooling systems, the sampling and maintenance of the tritium plant, the test blanket module system and the port plug test facilities, and from the laboratories. The operational and decommissioning waste inventories are summarized in Table 2. Note that the waste amount arising from the decommissioning of the machine is coming from the RPrS assessment. The consolidation of the decommissioning waste amounts is foreseen to be performed as the design progress. Radwaste will also be generated from the Test Blanket Module (TBM) Program during ITER operation and decommissioning. The associated radwaste management is detailed in [5].
3. Waste management routes As mentioned in [3], one of the main objectives of the waste management process, as emphasized by the 2006 Law [1] is to reduce waste radiotoxicity and amounts. The first way to do this is to limit waste production at source. The measures taken at design level are described in [3] (reduction of impurities in materials, use of shielding, design of the installations to facilitate the feasibility and effectiveness of the decontamination operations, possibility to segregate the different part of a component, etc.). The second way is to have a waste management route adapted to the waste characteristics and radiotoxicity. The waste routes are developed based on the available inventories. After segregation and sorting at source (smearing, dose rate, spectrometry . . .), the discarded parts are transferred to a dedicated facility as a function of their characteristics: • Solid MAVL waste is managed in the hot cell building, • Solid purely tritiated waste is managed in the tritium building and stored in the tritium building and hot cell building, • Solid FMA-VC waste is managed in the radwaste building, • Solid TFA waste is managed in the personnel access and control building, • Liquid radioactive effluent is managed in the radwaste building.
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Table 1 Waste classification of the main components at the end of ITER lifetime. Main component
Classification
Radioactive decay (hypothesis for classification)
Mass (t)
Blanket modules Divertor modules Vacuum vessel Toroidal field coils Poloidal field coils Central solenoid Cryostat
MAVL MAVL FMA-VC TFA TFA TFA TFA
No No No 50 years 50 years No 50 years
1530 650 5100 6010 1870 950 3500
Table 2 ITER operational and decommissioning waste inventories before processing.
Operational waste Process & housekeeping Component replacement Liquids effluents Deactivation and decommissioning Components Buildings
TFA
FMA-VC
Purely tritiated
MAVL
80 m3 /yr <10 t
200 m3 /yr 20–100 t 164 m3 /yr
110 t
<1 m3 /yr 1070 t
18,300 t 2350 t
8920 t 4300 t
390 t
2500 t
3.1. Solid MAVL waste Solid MAVL waste mainly arises from the replacement of invessel components (blanket, divertor). They are remotely removed from the vacuum vessel and transferred by cask to the hot cell building. During all the process, remote handling operations are foreseen to deal with the waste activity and radiation level. After dust cleaning, the discarded components are transferred to the waste treatment area where they are cut. Thermal desorption in an oven at 800 ◦ C is foreseen to reduce the tritium inventory before storage [6]. After sampling and characterization, the packaged waste is stored in the hot cell facility during the whole ITER operational lifetime. 3.2. Solid purely tritiated waste Solid purely tritiated waste arise from the maintenance of the systems of the tritium plant. Discarded components are sampled for characterization and packed in drums and interim stored in the tritium building before being sent to a storage area in the hot cell building. This area is dedicated to purely tritiated waste to avoid cross-contamination with activated and tritiated MAVL waste. The containers are stored in the hot cell facility during the whole ITER operational lifetime. 3.3. Solid FMA-VC waste Solid FMA-VC waste mainly arises from the maintenance of the neutron beam heating and diagnostic systems and from process and housekeeping. Such solid waste is transferred to the radwaste building where the following treatments are foreseen: cutting for large components, sorting, sampling and characterization, compaction and cementation. Once cemented and cured, the packages are stored in the radwaste building for 6 months before being exported. Contaminated oils, from the maintenance of the vacuum pumps and possible leaks of the water cooling circuit are also considered in this category. At this stage, it is assumed that 7 m3 of contaminated oils will be considered as waste each year. Current plans for the oil waste treatment process consist of a chemical pretreatment to precipitate radionuclide species as sludge, followed by two stage centrifuge separation to isolate the oil, water and sludge phases. The oil could be treated by off-site incineration facility if the
radiological acceptance criteria are satisfied. Otherwise they will be blocked in an improved cement or polymer matrix that still needs to be developed. The treatment of tritiated oils remains a challenge so effort is put to reduce as much as possible the production of such waste. Spent resins and filters from the water cooling circuit maintenance are stored in the buffer storage and feeding tanks, transferred to the cementation system, and stored for 6 months after curing. This will represent about 200 m3 of packaged waste every 6 months.
3.4. Solid TFA waste Solid TFA waste mainly arises from the maintenance of the vacuum systems and from process and housekeeping. Such waste is transferred in drums to the personnel and access control building where it is characterized and stored in packages for 6 months before being exported.
3.5. Liquid radioactive effluents Liquid radioactive effluents are transferred by pipe from the tokamak building to the radwaste building. The liquid waste treatment process mainly consists of filtration and evaporation. After the evaporation process, the evaporator concentrate is solidified using the cement solidification process. The distillate from the evaporation process passes through the ion exchange resin column to remove the possible carry-over of radionuclides. After sampling for radioactive content the processed effluent can be sent to discharge as industrial effluent or Water Detritiation System in Tritium Plant in accordance with the acceptance criteria. Solidified concentrates follow the same route as the solid FMAVC waste (see Section 3.3). About 3 m3 of packaged concentrates are expected every 6 months. Analysis of waste samples in laboratory will lead to the production of organic waste. The annual quantity produced is considered in the order of 120 l. This type of waste will require a specific evacuation route, such as off-site incineration, that is being studied.
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3.6. Waste disposal All the packaged waste, after temporary storage in ITER facilities (6 months for TFA and FMA-VC and for the whole ITER lifetime for MAVL and purely tritiated waste) will be sent to the Andra final disposal or to an interim storage for tritium decay if the waste can’t be directly accepted by the Andra because tritium content and outgassing [7]. 4. Main challenges related to the waste treatments 4.1. Cutting of mixed materials The cutting of large and thick components, in particular MAVL ones that are made of mixed materials (beryllium, tungsten, copper alloys and stainless steel) is one of the major challenges. As an example the dimensions of a divertor cassette is about 2.2 m × 3.3 m and it must be cut to fit into baskets of 1.5 m × 1.5 m × 1.45 m. The development of the process must consider the efficiency and reliability of the remote handled cutting but also the safety of the process. In particular, the dust, debris, gases arising from the cutting need to be collected as close as possible from the process to avoid contamination of the room. R&D is on-going on the cutting of large components. 4.2. Characterization Engineering studies are also being performed regarding the characterization of the waste. Gamma emitters will be measured by gamma spectrometry. For the measurement of specific nuclides that could not be measured by this technique, it is foreseen to use the correlation or scaling factor technique. The scaling factor method is based on developing a correlation between easily measurable gamma emitting nuclides (key nuclides) and “difficult to measure” nuclides. The activities of “difficult to measure” nuclides in waste packages are then estimated by measuring the gamma emitting nuclides and applying the scaling factors. It concerns mainly beta emitters and low energy gamma emitters that need to be characterized for Andra acceptation. For the divertor for example, it concerns Ca-45, V-49, Fe-55, Ni-59, Ni-63, Nb-91, Nb-93m, Mo-93, Sn-119m, Ta-179, W-181 and Re-184m. But these 2 methods can’t be used for tritium. As a matter of fact, as tritium is coming from activation but also permeation, correlation with other activated nuclides is not representative. Various methods are being developed for the measurement of tritium. Three methods have emerged as the most suitable for the waste characterization: • Sampling and radiological analysis (destructive method), • Calorimetry [8], • Measurement of He-3 [9]. The measurement of the He-3 is adapted only to technological waste (vinyl, cotton) and requires long waiting times to stabilize the outgassing and to be able to perform the measurement. This technique is therefore more suitable to old waste. The method is not suitable for metallic waste. Calorimetry is adapted to purely tritiated waste (detection limit of 1.8E + 14 Bq/drum). Its adaptation to waste with other gamma and beta emitters is possible in theory but has not yet been demonstrated. Sampling followed by radiological analysis (destructive method), despite the drawbacks related to sampling (representativeness, secondary waste,. . .), remains today the most adapted method for the measurement of tritium activity in ITER waste.
Outgassing measurements is also required by the Andra. The commonly used technique consists in having the package placed in a tight confinement enclosure whose internal atmosphere undergoes forced circulation and to measure the released gases. 4.3. Sampling The representativeness of the sampling is then essential and challenging considering the geometry of the waste. It concerns mainly MAVL waste and large FMA-VC waste. A specific sampling plan will be develop to ensure that the data obtained are always bounding values. The process put in place to ensure sample representativeness is based on the following principles: • Estimation of theoretical activity expected for the different discarded elements, taking into account experience feedback from components previously characterized, in order to identify areas likely to have the greatest activity. This estimation can be confirmed by gamma mapping to determine the “hot spots”. • Sampling by drilling or coring in areas with the highest dose rate. The sampling method must ensure that if there is loss of tritium due to outgassing during the process, these inventories will be accounted for. • Laboratory analysis to determine the spectrum of radionuclides present in the waste. There will be a laboratory for MAVL waste in the hot cell building and a laboratory for FMA-VC waste in the radwaste building. • Gamma spectrometry on the package to measure radionuclides. The determination of the “right” number of samples must also takes into account that sampling generates secondary waste that should be limited as much as possible. More generally, the strategy is initially based on a theoretical study that will be constantly confronted with actual measurements, with the aim to increase the knowledge on waste and ensure an envelope estimate of the specific activities in the waste. 4.4. Detritiation With regards to the detritiation technique to be used for MAVL solid waste, R&D is on-going to achieve an efficiency of 95% of tritium removal, while optimizing the process. Detritiation is to reduce further outgassing of the waste in the interim storage and disposal. Various detritiation techniques have been studied [10]: thermal desorption under atmospheric air, thermal desorption with isotopic exchange with gaseous hydrogen, heating with flame, vacuum melting, melting under static hydrogen atmosphere and under argon flow, etc. Of the available techniques, thermal desorption in an oven at 800 ◦ C in a flow of Argon gas (which may contain hydrogen not exceeding 4% in volume to prevent a hazardous atmosphere) is currently considered to be the most promising in terms of efficiency, ease of process and safety [11]. Studies are going to determine the optimized parameter, trying to remove the hydrogen while maintaining the efficiency of the process. The use of a detritiation system for the FMA-VC waste is also being analyzed. This could be developed in particular for process and housekeeping waste [12]. The major gain that would be expected is to reach a level of tritium remaining in the waste that would not require an interim storage before export to the Andra. 5. ITER facilities for waste management The hot cell complex groups the hot cell, the radwaste and the personnel access and control buildings. The complex design is progressing taking into account the updated inventories for the design
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of the waste processes and storage. The components participating to confinement or shielding including supporting function preventing the failure of the confinement and shielding are classified as protection important components (PIC/SIC) [2]. In the hot cell building, space is allocated to the in-vessel component interim storage before waste treatment, the waste treatment (cutting, sampling, detritiation with thermal desorption oven, packaging), the waste treatment services (including the gas treatment system of the detritiation oven), decontamination and control of the packages, interim storage, and import/export area. A laboratory is also foreseen. The MAVL waste processes in the hot cell building will be fully remotely operated and maintained. The hot cell also provides a dedicated area for the storage of purely tritiated waste. In the radwaste building, space is allocated to holding tanks (for radioactive liquid effluents, oils, spent resins and concentrates), monitoring tanks, oil separation process, buffer storage, characterization, sampling and sorting, cutting and dismantling workstation, compaction, cementation, curing, tritium outgassing measurement, temporary storage and export area. The operations in the radwaste building won’t be fully remote but ALARA (as low as reasonably achievable) must be followed to limit as much as possible the radiation exposure doses to the workers. In the personnel access and control building, space is allocated to the TFA buffer storage, characterization, laboratory and export area. The design of the storage areas shall ensure the possible recovery of the packages, throughout the lifetime of the storage (in normal operation or degraded), easily and safely within an appropriate period. This principle of recovery requires that the waste packages are monitored (traceability, radiological inventory, characterization, organized storage, etc.) and that the equipment necessary for their recovery is available. The management of damaged packages is made possible (re-packaging). 6. Conclusion Waste management routes are being developed for the different types of waste. The inventories of waste to be produced during
5
operation and decommissioning are frequently updated to take into account the latest progress of the design and the development of the maintenance schemes. Based on these inventories and on the specificity of waste (in particular the tritium content), processes are developed in the dedicated ITER buildings for waste management.
Acknowledgments The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.
References [1] Law 2006-739 of 28 June 2006 on the sustainable management of radioactive materials and waste. [2] Order of 7th of February 2012 laying down general rules for basic nuclear installations. [3] S. Rosanvallon, et al., ITER waste management, Fusion Eng. Des. 85 (2010) 1788–1791. [4] D. Canas, et al., Classification methodology for tritiated waste requiring interim storage, Fusion Sci. Technol. 67 (2015) 290–295. [5] J. van der Laan, Radwaste Management for the ITER TBM Program, this conference. [6] S. Rosanvallon, et al., Management of tritium in ITER waste, Fusion Sci. Technol. 60 (3) (2011) 855–860. [7] J. Pamela, et al., ITER tritiated waste management by the Host state and first lessons learned for fusion development, Fusion Eng. Des. 89 (2014) 2001–2007. [8] L. Dorr, et al., High resolution vacuum calorimeter, Fusion Sci. Technol. 48 (2005) 358–361. [9] D. Demange, et al., A review of the different methods to quantify tritium inside waste drums vial helium-3 mass spectrometric measurements, Fusion Sci. Technol. 48 (2005) 182–187. [10] A.N. Perevezentsev, et al., Detritiation studies for JET decommissioning, Fusion Eng. Des. 83 (2008) 1364–1367. [11] O. Gastaldi, et al., Optimisation of operating parameters for plasma facing components detritiation, Fusion Eng. Des. 84 (2009) 815–820. [12] K. Liger, et al., Preliminary results from a detritiation facility dedicated to soft housekeeping waste and tritium valorization, Fusion Eng. Des. 89 (2014) 2103–2107.
Please cite this article in press as: S. Rosanvallon, et al., Waste management plans for ITER, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.12.002
ARTICLE IN PRESS
FUSION-8372; No. of Pages 5
Fusion Engineering and Design xxx (2015) xxx–xxx
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Waste management plans for ITER S. Rosanvallon ∗ , D. Torcy, J.K. Chon, A. Dammann ITER Organization, CS 90 046, 13067 St Paul lez Durance Cedex, France
a r t i c l e
i n f o
Article history: Received 30 July 2015 Received in revised form 30 November 2015 Accepted 2 December 2015 Available online xxx Keywords: ITER Nuclear Waste Characterization Decommissioning
a b s t r a c t ITER will produce radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning. The waste concerns components that are activated by neutrons of energies up to 14 MeV, and are contaminated by activated corrosion products, activated dust and tritium. Even if the nuclear waste production will start only with the deuterium-deuterium phase, provisions have to be taken now with the design of the ITER facilities that will be used for the treatment and interim storage of the waste, in order to demonstrate that all the ITER waste will be safely manageable with the existing outlets. This demonstration also needs to be provided to the French regulator. © 2015 Elsevier B.V. All rights reserved.
1. Introduction Neutrons will be produced by deuterium–deuterium and deuterium–tritium plasmas in the ITER plasma chamber and will lead to the activation of the structures. In the same time, these components will be contaminated by activated corrosion products, activated dust and tritium. During maintenance operations, some in-vessel components are anticipated to be removed from the vacuum vessel moreover supporting systems such as tritium fuel cycle systems or heating systems are planned to be removed. These components, once discarded, as well as those dismantled during decommissioning, will be treated as radioactive waste. Provisions have to be taken at the design stage of the ITER facilities that will be used for the treatment and interim storage of the waste, in order to demonstrate that all the ITER waste will be safely manageable with the existing disposals. This demonstration also needs to be provided to the French regulator. The aim of this paper is to describe the waste origin, the waste inventories, the optimization process put in place to reduce the waste radiotoxicity and volumes, and the overall strategy from component removal to final disposal that can be foreseen at this stage of the project. The main challenges of the waste treatments are also highlighted. It has to be noted that radioactive waste management follows in France the framework of a law on the sustainable management of
∗ Corresponding author. Tel.: +33 442176786. E-mail address: [email protected] (S. Rosanvallon).
radioactive materials and waste established in June 2006 [1]. Waste management is a Protection Important Activity as per 2012 Order [2].
2. Waste characteristics 2.1. Radioactive content After a non-radioactive phase using hydrogen plasmas, ITER, which is the Nuclear Facility INB-174, will operate with deuterium plasmas (D–D) and deuterium–tritium plasmas (D–T). The activation of the structures will start with the 2.5 MeV neutrons resulting from the D–D reactions and will continue with the 14 MeV neutrons arising from the D–T reaction. In addition to activation, components may be contaminated by tritium, activated and tritiated dust and activated corrosion products (ACPs). In fact, as soon as tritium is used as fuel for the fusion reaction, in-vessel components will be contaminated by tritium adsorption and permeation. Fuelling circuit components (fuelling system and tritium plant) will not be activated but will be tritiated due to tritium permeation. Moreover, the plasma interacts with the plasma facing components during steady-state operation and transients (mainly disruptions) leading to the creation of dust that will have the same radiological characteristics as the first few microns of the plasma facing components and will be thus both activated and tritiated. Furthermore, water circulating in the heat transfer systems will corrode the pipes, leading to activated corrosion products in the fluids and pipe circuits.
http://dx.doi.org/10.1016/j.fusengdes.2015.12.002 0920-3796/© 2015 Elsevier B.V. All rights reserved.
Please cite this article in press as: S. Rosanvallon, et al., Waste management plans for ITER, Fusion Eng. Des. (2015), http://dx.doi.org/10.1016/j.fusengdes.2015.12.002
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2.2. Radioactive waste As mentioned in [3], there is no exemption and/or clearance in France. The regulations impose that the nuclear operator defines a waste zoning of the facilities. It consists in defining zones producing radioactive waste and those producing non-radioactive waste. These zones are defined based on an analysis of the normal operations foreseen room by room and the potential exposure to activation and/or contamination. All the waste arising from a zone producing radioactive waste will have to be managed as radioactive waste as a function of the Andra (French national radioactive waste management agency) classification for radioactive waste. The operation and decommissioning of ITER will lead to gaseous effluents, liquids effluents and solid waste. The radioactive gaseous effluents are released to the environment via the ventilation and/or detritiation systems and through a monitored exhaust of the Tokamak Complex. They are not treated in this paper. The radioactive liquids effluents consists in effluents with radioactivity levels exceeding the limit values for discharge into the industrial sewage system and that need to be further processed to be considered as waste. The solid radioactive waste arises from day-to-day operation, maintenance and as a result of component replacement from radioactive waste zones. After treatment and solidification, liquid effluents lead also to solid waste. The solid waste classification is managed by the Andra. ITER radioactive waste can be classified into three categories based on its activity and radio-toxicity (nota: French abbreviations are used since they correspond to specific acceptation criteria): • TFA waste (French abbreviation for “Très Faible Activité” meaning very low-level), • FMA-VC waste (French abbreviation for “Faible et Moyenne Activité à Vie Courte” meaning low and intermediate-level short-lived) or type A waste, • MA-VL waste (French abbreviation for “Moyenne Activité à Vie Longue” meaning intermediate-level long-lived) or type-B waste. No HAVL waste (French abbreviation for “Haute Activité à Vie Longue” meaning high-level long-lived waste) is generated by the ITER facility. In addition to the ANDRA categories, because of the use of tritium as fuel in ITER, purely tritiated waste arising from the tritium plant for example, is also considered. This consideration is to enable a specific routing for this type of waste to avoid crosscontamination with waste that is activated and tritiated during interim storage. After radioactive decay, it is anticipated that this waste will be sent to the FMA-VC storage centre. The criteria for waste classification are given in [4]. 2.3. Inventories During the licensing process, the waste inventories are assessed to demonstrate that all the waste produced by the facility are manageable from the production (on-site treatments and interim storage) to the off-site treatments, if any, and final disposal. A first assessment has been prepared for the Preliminary Safety Report (RPrS) that has been submitted to the French Regulator in 2011. This assessment was established considering an inventory based on the design and the maintenance plans as they were standing and the classification was based on 1-D activation calculations. Since then, 3-D activation calculations have been carried out to reflect the latest updates on the materials used, the design of the machine and the planned maintenance operations. For the waste classification, tritium due to sources of trapping such as
implantation, recombination or permeation has been added to the tritium due to the activation. The classification of the main components at the end of the ITER lifetime is presented in Table 1. In some cases, the classification is given after a radioactive decay time after plasma shutdown. It concerns only components that are to be discarded during decommissioning. With these new calculations, the inner wall of the vacuum vessel is slightly above the acceptance criteria for FMA-VC waste. Nevertheless, when the vacuum vessel is considered globally (inner wall, shielding and outer wall all together), the classification remains FMA-VC as presented in Table 1. This considers also the presence of tritium in the waste. These results need to be further assessed as they could impact the decommissioning. Decommissioning scenarios were also proposed in the RPrS. They are being updated and based on the scenario that will be considered, it may be needed to reconsider this classification. A review of the ITER operational waste has also been performed in 2014. The operational waste consists in: • Process and housekeeping waste (gloves, papers, ion exchange resins, filters, oils, etc.) as in all nuclear facility, • Component replacement linked to the experimental program on plasma facing materials and to the maintenance of supporting systems such as tritium fuel cycle systems or heating systems, • Liquid effluents from the maintenance of the cooling systems, the sampling and maintenance of the tritium plant, the test blanket module system and the port plug test facilities, and from the laboratories. The operational and decommissioning waste inventories are summarized in Table 2. Note that the waste amount arising from the decommissioning of the machine is coming from the RPrS assessment. The consolidation of the decommissioning waste amounts is foreseen to be performed as the design progress. Radwaste will also be generated from the Test Blanket Module (TBM) Program during ITER operation and decommissioning. The associated radwaste management is detailed in [5].
3. Waste management routes As mentioned in [3], one of the main objectives of the waste management process, as emphasized by the 2006 Law [1] is to reduce waste radiotoxicity and amounts. The first way to do this is to limit waste production at source. The measures taken at design level are described in [3] (reduction of impurities in materials, use of shielding, design of the installations to facilitate the feasibility and effectiveness of the decontamination operations, possibility to segregate the different part of a component, etc.). The second way is to have a waste management route adapted to the waste characteristics and radiotoxicity. The waste routes are developed based on the available inventories. After segregation and sorting at source (smearing, dose rate, spectrometry . . .), the discarded parts are transferred to a dedicated facility as a function of their characteristics: • Solid MAVL waste is managed in the hot cell building, • Solid purely tritiated waste is managed in the tritium building and stored in the tritium building and hot cell building, • Solid FMA-VC waste is managed in the radwaste building, • Solid TFA waste is managed in the personnel access and control building, • Liquid radioactive effluent is managed in the radwaste building.
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Table 1 Waste classification of the main components at the end of ITER lifetime. Main component
Classification
Radioactive decay (hypothesis for classification)
Mass (t)
Blanket modules Divertor modules Vacuum vessel Toroidal field coils Poloidal field coils Central solenoid Cryostat
MAVL MAVL FMA-VC TFA TFA TFA TFA
No No No 50 years 50 years No 50 years
1530 650 5100 6010 1870 950 3500
Table 2 ITER operational and decommissioning waste inventories before processing.
Operational waste Process & housekeeping Component replacement Liquids effluents Deactivation and decommissioning Components Buildings
TFA
FMA-VC
Purely tritiated
MAVL
80 m3 /yr <10 t
200 m3 /yr 20–100 t 164 m3 /yr
110 t
<1 m3 /yr 1070 t
18,300 t 2350 t
8920 t 4300 t
390 t
2500 t
3.1. Solid MAVL waste Solid MAVL waste mainly arises from the replacement of invessel components (blanket, divertor). They are remotely removed from the vacuum vessel and transferred by cask to the hot cell building. During all the process, remote handling operations are foreseen to deal with the waste activity and radiation level. After dust cleaning, the discarded components are transferred to the waste treatment area where they are cut. Thermal desorption in an oven at 800 ◦ C is foreseen to reduce the tritium inventory before storage [6]. After sampling and characterization, the packaged waste is stored in the hot cell facility during the whole ITER operational lifetime. 3.2. Solid purely tritiated waste Solid purely tritiated waste arise from the maintenance of the systems of the tritium plant. Discarded components are sampled for characterization and packed in drums and interim stored in the tritium building before being sent to a storage area in the hot cell building. This area is dedicated to purely tritiated waste to avoid cross-contamination with activated and tritiated MAVL waste. The containers are stored in the hot cell facility during the whole ITER operational lifetime. 3.3. Solid FMA-VC waste Solid FMA-VC waste mainly arises from the maintenance of the neutron beam heating and diagnostic systems and from process and housekeeping. Such solid waste is transferred to the radwaste building where the following treatments are foreseen: cutting for large components, sorting, sampling and characterization, compaction and cementation. Once cemented and cured, the packages are stored in the radwaste building for 6 months before being exported. Contaminated oils, from the maintenance of the vacuum pumps and possible leaks of the water cooling circuit are also considered in this category. At this stage, it is assumed that 7 m3 of contaminated oils will be considered as waste each year. Current plans for the oil waste treatment process consist of a chemical pretreatment to precipitate radionuclide species as sludge, followed by two stage centrifuge separation to isolate the oil, water and sludge phases. The oil could be treated by off-site incineration facility if the
radiological acceptance criteria are satisfied. Otherwise they will be blocked in an improved cement or polymer matrix that still needs to be developed. The treatment of tritiated oils remains a challenge so effort is put to reduce as much as possible the production of such waste. Spent resins and filters from the water cooling circuit maintenance are stored in the buffer storage and feeding tanks, transferred to the cementation system, and stored for 6 months after curing. This will represent about 200 m3 of packaged waste every 6 months.
3.4. Solid TFA waste Solid TFA waste mainly arises from the maintenance of the vacuum systems and from process and housekeeping. Such waste is transferred in drums to the personnel and access control building where it is characterized and stored in packages for 6 months before being exported.
3.5. Liquid radioactive effluents Liquid radioactive effluents are transferred by pipe from the tokamak building to the radwaste building. The liquid waste treatment process mainly consists of filtration and evaporation. After the evaporation process, the evaporator concentrate is solidified using the cement solidification process. The distillate from the evaporation process passes through the ion exchange resin column to remove the possible carry-over of radionuclides. After sampling for radioactive content the processed effluent can be sent to discharge as industrial effluent or Water Detritiation System in Tritium Plant in accordance with the acceptance criteria. Solidified concentrates follow the same route as the solid FMAVC waste (see Section 3.3). About 3 m3 of packaged concentrates are expected every 6 months. Analysis of waste samples in laboratory will lead to the production of organic waste. The annual quantity produced is considered in the order of 120 l. This type of waste will require a specific evacuation route, such as off-site incineration, that is being studied.
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3.6. Waste disposal All the packaged waste, after temporary storage in ITER facilities (6 months for TFA and FMA-VC and for the whole ITER lifetime for MAVL and purely tritiated waste) will be sent to the Andra final disposal or to an interim storage for tritium decay if the waste can’t be directly accepted by the Andra because tritium content and outgassing [7]. 4. Main challenges related to the waste treatments 4.1. Cutting of mixed materials The cutting of large and thick components, in particular MAVL ones that are made of mixed materials (beryllium, tungsten, copper alloys and stainless steel) is one of the major challenges. As an example the dimensions of a divertor cassette is about 2.2 m × 3.3 m and it must be cut to fit into baskets of 1.5 m × 1.5 m × 1.45 m. The development of the process must consider the efficiency and reliability of the remote handled cutting but also the safety of the process. In particular, the dust, debris, gases arising from the cutting need to be collected as close as possible from the process to avoid contamination of the room. R&D is on-going on the cutting of large components. 4.2. Characterization Engineering studies are also being performed regarding the characterization of the waste. Gamma emitters will be measured by gamma spectrometry. For the measurement of specific nuclides that could not be measured by this technique, it is foreseen to use the correlation or scaling factor technique. The scaling factor method is based on developing a correlation between easily measurable gamma emitting nuclides (key nuclides) and “difficult to measure” nuclides. The activities of “difficult to measure” nuclides in waste packages are then estimated by measuring the gamma emitting nuclides and applying the scaling factors. It concerns mainly beta emitters and low energy gamma emitters that need to be characterized for Andra acceptation. For the divertor for example, it concerns Ca-45, V-49, Fe-55, Ni-59, Ni-63, Nb-91, Nb-93m, Mo-93, Sn-119m, Ta-179, W-181 and Re-184m. But these 2 methods can’t be used for tritium. As a matter of fact, as tritium is coming from activation but also permeation, correlation with other activated nuclides is not representative. Various methods are being developed for the measurement of tritium. Three methods have emerged as the most suitable for the waste characterization: • Sampling and radiological analysis (destructive method), • Calorimetry [8], • Measurement of He-3 [9]. The measurement of the He-3 is adapted only to technological waste (vinyl, cotton) and requires long waiting times to stabilize the outgassing and to be able to perform the measurement. This technique is therefore more suitable to old waste. The method is not suitable for metallic waste. Calorimetry is adapted to purely tritiated waste (detection limit of 1.8E + 14 Bq/drum). Its adaptation to waste with other gamma and beta emitters is possible in theory but has not yet been demonstrated. Sampling followed by radiological analysis (destructive method), despite the drawbacks related to sampling (representativeness, secondary waste,. . .), remains today the most adapted method for the measurement of tritium activity in ITER waste.
Outgassing measurements is also required by the Andra. The commonly used technique consists in having the package placed in a tight confinement enclosure whose internal atmosphere undergoes forced circulation and to measure the released gases. 4.3. Sampling The representativeness of the sampling is then essential and challenging considering the geometry of the waste. It concerns mainly MAVL waste and large FMA-VC waste. A specific sampling plan will be develop to ensure that the data obtained are always bounding values. The process put in place to ensure sample representativeness is based on the following principles: • Estimation of theoretical activity expected for the different discarded elements, taking into account experience feedback from components previously characterized, in order to identify areas likely to have the greatest activity. This estimation can be confirmed by gamma mapping to determine the “hot spots”. • Sampling by drilling or coring in areas with the highest dose rate. The sampling method must ensure that if there is loss of tritium due to outgassing during the process, these inventories will be accounted for. • Laboratory analysis to determine the spectrum of radionuclides present in the waste. There will be a laboratory for MAVL waste in the hot cell building and a laboratory for FMA-VC waste in the radwaste building. • Gamma spectrometry on the package to measure radionuclides. The determination of the “right” number of samples must also takes into account that sampling generates secondary waste that should be limited as much as possible. More generally, the strategy is initially based on a theoretical study that will be constantly confronted with actual measurements, with the aim to increase the knowledge on waste and ensure an envelope estimate of the specific activities in the waste. 4.4. Detritiation With regards to the detritiation technique to be used for MAVL solid waste, R&D is on-going to achieve an efficiency of 95% of tritium removal, while optimizing the process. Detritiation is to reduce further outgassing of the waste in the interim storage and disposal. Various detritiation techniques have been studied [10]: thermal desorption under atmospheric air, thermal desorption with isotopic exchange with gaseous hydrogen, heating with flame, vacuum melting, melting under static hydrogen atmosphere and under argon flow, etc. Of the available techniques, thermal desorption in an oven at 800 ◦ C in a flow of Argon gas (which may contain hydrogen not exceeding 4% in volume to prevent a hazardous atmosphere) is currently considered to be the most promising in terms of efficiency, ease of process and safety [11]. Studies are going to determine the optimized parameter, trying to remove the hydrogen while maintaining the efficiency of the process. The use of a detritiation system for the FMA-VC waste is also being analyzed. This could be developed in particular for process and housekeeping waste [12]. The major gain that would be expected is to reach a level of tritium remaining in the waste that would not require an interim storage before export to the Andra. 5. ITER facilities for waste management The hot cell complex groups the hot cell, the radwaste and the personnel access and control buildings. The complex design is progressing taking into account the updated inventories for the design
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of the waste processes and storage. The components participating to confinement or shielding including supporting function preventing the failure of the confinement and shielding are classified as protection important components (PIC/SIC) [2]. In the hot cell building, space is allocated to the in-vessel component interim storage before waste treatment, the waste treatment (cutting, sampling, detritiation with thermal desorption oven, packaging), the waste treatment services (including the gas treatment system of the detritiation oven), decontamination and control of the packages, interim storage, and import/export area. A laboratory is also foreseen. The MAVL waste processes in the hot cell building will be fully remotely operated and maintained. The hot cell also provides a dedicated area for the storage of purely tritiated waste. In the radwaste building, space is allocated to holding tanks (for radioactive liquid effluents, oils, spent resins and concentrates), monitoring tanks, oil separation process, buffer storage, characterization, sampling and sorting, cutting and dismantling workstation, compaction, cementation, curing, tritium outgassing measurement, temporary storage and export area. The operations in the radwaste building won’t be fully remote but ALARA (as low as reasonably achievable) must be followed to limit as much as possible the radiation exposure doses to the workers. In the personnel access and control building, space is allocated to the TFA buffer storage, characterization, laboratory and export area. The design of the storage areas shall ensure the possible recovery of the packages, throughout the lifetime of the storage (in normal operation or degraded), easily and safely within an appropriate period. This principle of recovery requires that the waste packages are monitored (traceability, radiological inventory, characterization, organized storage, etc.) and that the equipment necessary for their recovery is available. The management of damaged packages is made possible (re-packaging). 6. Conclusion Waste management routes are being developed for the different types of waste. The inventories of waste to be produced during
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operation and decommissioning are frequently updated to take into account the latest progress of the design and the development of the maintenance schemes. Based on these inventories and on the specificity of waste (in particular the tritium content), processes are developed in the dedicated ITER buildings for waste management.
Acknowledgments The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.
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