1. Some technological problems of fusion materials management Boris N. Kolbasov a *, Laila El-Guebaly b, Vladimir I. Khripunov a, Youji Someya c, Kenji Tobita c, Massimo Zucchetti d a National Research Center “Kurchatov Institute”, Moscow, Russia b University of Wisconsin-Madison, Madison, Wisconsin, USA c Japan Atomic Energy Agency, Japan d EURATOM/ENEA Fusion Association, Politecnico di Torino, Torino, Italy ISFNT-11, Barcelona, 2013

2. 2 Content of the report 1. Strategy for handling fusion activated materials. 2. Impurities in fusion materials. 3. Clearance possibility for components of advanced fusion power cores (FPCs). 4. Radioactivity build-up by multiple reuse of divertor made of W-La 2 O 3. 5.C-14 generation in FPCs and its effect on material disposal and clearance. 6. Waste management scenario under replacement of blanket and divertor.

3. ISFNT-11, Barcelona, 20133 1. Strategy for handling activated fusion materials This international collaborative study on management of fusion radioactive materials has been carried out within the framework of the International Energy Agency Program on Environmental, Safety and Economic Aspects of Fusion Power to examine the back-end of the materials cycle. The strategy for handling fusion activated materials calls for three potential schemes: clearance, recycling and disposal. There is a growing international effort to avoid geologic disposal, for fusion in particular. Clearance and recycling offer more environmentally attractive options through the reuse of activated materials.

4. ISFNT-11, Barcelona, 20134 Possibility of clearance and recycling All the fusion component materials could potentially be recycled [1-4] providing that advanced radiation- resistant remote handling equipment is capable of of handling doses ≥10 kSv/hr and could also be adapted for fusion use (component size, weight).

5. ISFNT-11, Barcelona, 20135 Beryllium resources and fusion needs Be is a scarce metal. The world natural resources of Be are somewhat higher than 80 000 t. About 65% of them are in USA. Be production in the world in 2012 was ~230 t: USA (88%), China (9%) and Mozambique (1%) [1]. Amount of Be in Russian DEMO-S FPR (2.4 GW) —215 t; in SEAFP PM3 (3 GW) —865 t; in PPCS-B (3.6 GW) —560 t. Lifetime of Be in FPRs is expected to be ~ 5 years due to swelling. Thus, recycling of Be irradiated in FPRs seems inevitable. [1] US Geological Survey, Be Statistics and Information, March 2013.

6. ISFNT-11, Barcelona, 20136 2. Impurities in fusion materials. Two assumptions were made in this report: for the demonstration reactors and FPPs of the first generation we considered present-day materials with minimum impurity contents reached by the industry. for the advanced FPPs of the second generation, e.g. for the ARIES designs, the materials with the lowest impurity content ever achieved in large-scale fabrication practices or restricted by technological requirements were considered. Unfortunately, present-day materials often do not meet FPP requirements, in particular with respect to impurities concentration (e.g. U in beryllium, Nb, Co, Mo, Ag, N in beryllium and in other materials).

7. ISFNT-11, Barcelona, 20137 3. Clearance possibility for components of advanced fusion power cores (FPCs). Plasma facing components (divertor and blanket) normally contain high radioactivity and are not clearable. As clearance of sizeable components (such as biological shield, cryostat, vacuum vessel, and some constituents of magnets) is highly desirable, we identified the source of radioisotopes that hinder the clearance of these components and investigated the impact of impurity control.

8. ISFNT-11, Barcelona, 20138 ARIES-ACT-1 isometric, outboard (OB) radial build, and volumes of fusion power core components

9. ISFNT-11, Barcelona, 20139 Clearance indices of fully compacted outboard components of ARIES-ACT-1. All the ARIES-ACT-1 components operate for ~50 years with 85% availability, except the first wall (FW), blanket-I and divertor (replaceable every ~5 years). The neutron wall loading averages 3.4 MW/m2 at the outboard FW. Only the cryostat and bioshield are clearable (for concrete of bioshield at ~1 year and for cryostat at ~70 years after plant shutdown).

10. ISFNT-11, Barcelona, 201310 Though the magnet is well protected by the blanket, VV and shield, it is not clearable even after 100 years of storage. The CI of Nb3Sn conductor is the dominant and only the Cu stabilizer could be cleared shortly after plant shutdown. Clearance indices of outboard magnet constituents.

11. ISFNT-11, Barcelona, 201311 Contribution of different radioisotopes to clearance indices of VV at 100 years after shutdown The main sources of radioisotopes contributing to OB VV Ci equal to 770 after 100 years of cooling are impurities (0.5 wppm Nb, 0.02 wppm Eu, and 8 wppm Co), except for 14 C generated from the 0.1wt% carbon alloying element. However, a VV made of pure iron would have a CI of ~20 at 100 years of storage (from 60 Co generated by Fe).

12. ISFNT-11, Barcelona, 201312 Impossibility of VV clearance after 100-yr storage This means no matter how serious an attempt is to control all the impurities or alter the alloying elements of the 3Cr-3WV steel, the VV (and shield) will never qualify for clearance, but remains recyclable, however. Recycling is the only viable option to avoid disposing the fusion power core activated materials and to minimize the radioactive waste volume assigned for geological repositories.

13. ISFNT-11, Barcelona, 201313 4. Radioactivity build-up by multiple reuse of divertor made of W-La 2 O 3. Recycling contact dose rate of divertor with W―La2O3 alloy after 1, 2, 5, and 10 cycles. In ARIES a divertor cycle consists of 3.4 FPY of operation followed by a 3 year of cooling, refabrication and inspection. After two cycles, the divertor can be handled after 2 days of cooling (using advanced remote handling system). After 10 cycles, ~4 months must pass before the divertor can be handled. This is well within the allowable 3 years before insertion into the FPP. Such a build-up of radioactivity resulting from the multiple reuse of the divertor is acceptable.

14. ISFNT-11, Barcelona, 201314 5.C-14 generation in FPPs The 14 C generation rate is determined by material composition of a blanket and neutron spectrum. Production of 14 C in steel, Be and V alloys depends mainly on the N impurity. In water, concrete and ceramic tritium breeders it is determined by 17 O. In СFС and SiC composites noticeable contribution gives 13 C. Abundance of 14 N, 17 O and 13 C in the corresponding natural elements is 99.6, 0.038 and 1.1%, respectively.

15. ISFNT-11, Barcelona, 201315 C-14 effect on disposal of some fusion materials At 12 MW·yr/m 2 : eurofer-97 BeV-5Ti-5Cr Concentration of N, wppm 300 280 150 Specific activity of 14 C, MBq/kg 400 380 200 Limits for near-surface disposal, MBq/kg US NRC 380 1600 500 Rokkasho-mura, Japan 100 100 100 El-Cabri, Spain 3.7 3.7 3.7

16. ISFNT-11, Barcelona, 201316 C-14 effect on disposal of SiC f /SiC composite At parameters of the reactor ARIES-ACT-1 given earlier for blanket lifetime of 5 years, requirements of US NRC for near surface disposal of SiC f /SiC composite will be met, if N concentration in SiC does not exceed 80 wppm. N concentration in present-day SiC f /SiC composite exceeds 1000 wppm. Therefore 14 C production rate in the blanket with such a composite is higher than in the blankets of the 1 st generation FPPs.

17. ISFNT-11, Barcelona, 201317 C-14 effect on recycling of materials The main radionuclides that impede (or may impede) recycling procedures are tritium and 14 C. If tritium can be removed before recycling, removal of 14 C before recycling is practically impossible. At different recycling operations carbon easily reacts with oxygen and hydrogen forming CO 2 and CH 4 and in smaller degree CO and C 2 H 6. Unfortunately, presence of oxygen and hydrogen during such operations is usually unavoidable. Therefore carbon trapping, apparently, should be foreseen at designing recycling technology. The confident answer requires knowledge of applied recycling procedures.

18. ISFNT-11, Barcelona, 201318 C-14 effect on clearance of bioshield concrete Estimation of 14 C impact on clearance of bioshield concrete shows that, since oxygen content in typical concrete is ~48 wt.%, specific activity of 14 C resulting from neutron interaction with concrete oxygen at neutron flux of 1∙10 7 n/(cm 2 ∙s) during 42.5 FPYs will exceed IAEA clearance limit for 14 C equal to 1 Bq/g. In ITER average neutron flux on the bioshield is about 1∙10 8 n/(cm 2 ∙s). In neutron hot spots it is up to 1∙10 9 n/(cm 2 ∙s). neutron activation of other impurities (Eu, Co and Cs) in neutron hot spots may aggravate the problem.

19. ISFNT-11, Barcelona, 201319 6. Waste management scenario under replacement of blanket and divertor. Under radioactive component replacement, the assembly of blanket or divertor modules should be removed from the reactor to minimize remote handling inside the VV and to attain reasonable plant availability. In the hot cell the modules will be separated from the back plate. Since the back plate made of reduced activation ferritic/martensitic (RAFM) steel is reused, its temperature should be kept below 550 ºC to maintain its structural strength. Maintenance scheme for the Slim CS DEMO fusion reactor

20. ISFNT-11, Barcelona, 201320 Requirements to RAFM steel temperature maintaining At natural air convection maximum back plate temperature exceeds 900 o C. Therefore the sector assembly in the hot cell needs to be forced cooled during 6 months. To this end, the cooling water systems of the sector assembly and the hot cell should be connected. In this case the temperature of the assembly can be maintained at 40-100 o C, but the connection with re-welding and inspection must be performed during a few hours to avoid the assembly overheating.

21. ISFNT-11, Barcelona, 201321 Requirements to mortar temperature maintaining The structural material (RAFM) of the blanket and divertor is subject to disposal. It should be crushed into small pieces to reduce its volume and required storage space. Then it is packaged in 1.6-m cubic containers with mortar for proper disposal. The decay heat must be removed by natural convection to keep the temperature below 65 ºC for preventing water evaporation from the mortar. The RAFM is kept in the interim storage during 12 years until the required temperature conditions for mortar are ensured and then is disposed of.

22. ISFNT-11, Barcelona, 201322 Conclusions - 1 The activity of fusion materials, irradiation doses from them, possibility of their clearance, conditions for their recycling and disposal are determined to a great extent (and quite often mainly) by amount of impurities. Designing reactors to be constructed in near future, we prefer to consider materials that the industry can supply without long development and additional cost. In particular it relates to amount of impurities. However the reactors that should become a base for future fusion power engineering, require materials with the lowest attainable and technologically permissible impurity content, as e.g. assumed in the ARIES designs.

23. ISFNT-11, Barcelona, 201323 Conclusions - 2 Under conditions assumed for FPP ARIES-ACT-1, only the cryostat and bioshield are clearable. Among magnet constituents only the Cu stabilizer could be cleared shortly after plant shutdown. A VV made of pure iron is not even clearable at 100 years of storage (due to 60 Co generated by iron. Recycling is the only viable option to avoid disposing the activated PFC materials and to minimize the radioactive waste volume assigned for geological repositories.

24. ISFNT-11, Barcelona, 201324 Conclusions - 3 Production of 14 C may cause problems for radioactive waste disposal and clearance of such FPP components as bioshield. The cooling time for decay heat reduction of recyclable radioactive components before their withdrawal from the FPPs may depend on the time required for connection of the cooling water systems of the recyclable component and the hot cell.