journal of ELSEVIER
nuclear materials
Journal of Nuclear Materials 233-237 (1996) 1500-1504
Recycling and clearance of fusion activated waste Paolo Rocco
a,
* , Massimo Zucchetti
b
• Safety Technology Institute, Joint Research Centre, European Commission, Ispra (VA), Italy b Department of Energetics, Polytechnic of Turin, Turin, Italy
Abstract
The irradiation conditions of ITER (International Thermonuclear Experimental Reactor) are taken as reference to analyze recycling (re-use of the waste material after suppression of noxious radionuclides) and clearance (declassification to non-active waste). Recycling is assessed, assuming limits for the surface dose rates of the waste. If alternative materials (reduced activation ferritic steel or V-Ti alloys) are used, the in-vessel components can be almost completely recycled. A vacuum vessel made of INCONEL and reduced activation steel can also be recycled. Clearance is assessed for the out-of-vessel components, using weighted averages of the clearance levels for single radionuclides proposed by recent IAEA studies. All the components of the outboard zones can be declassified to non-active waste after decay times from a few years to 50 y. The higher irradiation conditions of the inboard zone require the recycling of the coils and the use of reduced activation steel to allow declassification of the coil casings.
1. Introduction
Beside the achievement of scientific and technological goals, the success of fusion as a viable energy source for the next century requires a large effort aiming at reducing as much as possible the amount of radioactive waste and its long-term radioactivity. This result is required both for public acceptance and the technical aspects of radioactive waste management. A survey on the perception of risk in Japan and in the US associated to 30 activities, substances, and technologies shows that 'radioactive waste' ranges among the greatest dreads in both countries and arising almost the same fear as 'nuclear wars' and 'nuclear accidents' [l]. Furthermore, even if safety analyses demonstrate that the potential hazard of waste from fusion will be lower by orders of magnitude than that of the fission waste, it is hardly credible that the public perception will differ for the two kinds of waste. Concerning waste management aspects, materials under development at present for fusion applications have compositions optimized to reduce the long-term radioactivity,
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so as to allow simple waste management procedures, i.e. near-surface disposal. On the other hand, the national regulations on radioactive waste management may hamper the simple disposal procedures envisaged. Three examples of countries which have shown interest to host ITER, thus showing to be 'fusion-concerned' are given hereunder. (1) In the US, fusion materials should have radioactivity levels of long-lived nuclides within the limits for Class C, the most restrictive of the three categories of low-level waste (LLW) defined in US IOCFR61 [2], for which a 'shallow land burial' (SLB) is envisaged. These rules were originally developed for LL W arising from fission and were extended to fusion-specific radioisotopes [3]. The suitability of this extension however has raised some criticism [4] which, if taken into account, should cause substantial reductions of these limits. Thus, it is possible that SLB will not be applicable to fusion waste, which will either pertain to the last LL W category, the 'greater than class C' (GTCC), for which a disposal to a depth greater (but unspecified) than SLB is envisaged, or to high level waste (HLW), to be disposed (in the future) to geological depth [5]. (2) In France, shallow land burial is limited to beta and gamma emitters with half lives of less than 30 y, assuming an institutional control of the dump of 300 y and an
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P. Rocco, M. Zucchetti /Journal of Nuclear Materials 233-237 (1996) 1500-1504
unlimited access after this time span [5]. All other waste goes to a geological disposal. Fusion waste will enter in this category. (3) In Germany, all radioactive waste must be disposed in geological repositories: (a) as non-heat generating waste (NHGW) if the waste will lead to temperature increase in the host rock of less than 3 K. The surface dose should be less than 2 mSv /h to allow unshielded handling, or (b) as heat generating waste (HGW) where the host rock can have a temperature rise up to hundreds of degrees. Remote handling is envisaged for this waste, so that high surface dose rates are allowed.
2. Proposed management strategy for fusion waste Taking into account the aspects of public perception and waste management regulations, a waste management strategy is examined here where: (a) After pre-treatment, fusion waste is subjected to an interim storage at the plant site. This period should last 50 y for the most active components. It is expected that after this time lapse the activity levels of the reduced-activation materials used in commercial fusion reactors will be strongly reduced. Then, options (b) or (c) should be adopted. (b) Waste with high residual radioactivity is partially (or totally) recycled, i.e., is reprocessed to eliminate the noxious radionuclides and is (eventually) re-used in the nuclear industry. (c) Waste with radioactivity levels comparable to natural substances are 'cleared', i.e., declassified to non-active waste, with less stringent requirements for dumping. This strategy is evidently suitable for fusion power reactors, made with materials with reduced long-term radioactivity. ITER however can be a useful test-bed for the application of these concepts to some of its materials/components and for extrapolation to alternative materials showing reduced activation.
3. Recycling Waste reprocessing depends on low enough values of the surface dose rates and on the capability to extract the noxious radionuclides by chemical-physical processes. A recent surface dose rate-related classification [6] assumes IO, 100 µSv /h, l and 10 mSv /h as limits for hands-on processing, modified hands-on processing (minimizing exposure time, shadow shield, etc.), quasi-remote processing (shielded cab) and remote-processing (hot cells and remote handling) respectively. A classification proposed by the authors [7] takes 25 µSv /h as limits for hands-on recycling whereas 2 and 20 mSv /h are limits for
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'limited' and 'extended' recycling scenarios. Surface dose rate limits derived from these studies are adopted here. The reduction of the radioactivity levels depends on the feasibility of separating the relevant radionuclides from the waste. A study on V-15Cr-5Ti [6] shows that the main responsible of long-term activation, the short-lived K-42, can be effectively eliminated by melting the irradiated alloy. Recent assessments show that copper can be reprocessed also [8).
4. Clearance Clearance means: "the release of radiation sources from regulatory control" [9]. The declassification of fusion activated waste to nonactive waste has already been analyzed by the authors in the frame of the SEAFP assessments [10) which concerned the safety and environmental aspects of power fusion reactors. Quantitative results on the sorting of the activated waste from this reactor are reported in [11). About 2900 tons, i.e. only 5% of the activated waste amount, pertaining to the outer zones can be 'cleared' after an interim storage of 50 y at the reactor site. This unsatisfactory result is mainly due to the use of 'conventional' materials not optimized from the point of view of long-term radioactivity for the outer part of the SEAFP reactor. Furthermore, the 'clearance' (or exemption) limit, was taken from the radioactivity limit of the british regulations for the very low level waste [12), i.e., 400 Bq/kg. It is a very low value, also compared with the natural radioactivity of many substances. Fertilizers and bricks for instance may exceed 5000 and 1000 Bq/kg. Unconditional clearance levels for radionuclides in solid materials are given in [9], which is yet in a draft form but will be hopefully issued as an IAEA recommendation, possibly with some modified figures if deemed necessary. Most of these levels are derived from the categorization of results of safety analyses where the annual individual dose for 'likely' and 'unlikely' scenarios of accidents to the waste repository is limited to 10 and 100 µSv per y, respectively. A fitting formula can be used for radionuclides not included in the categorization. The clearance level is given by: . . { Mmmum
I ALiinh ALiing } --- --EY + 0.1 E13, 1000 , 100000
(Bq/gorBq/cm 2 ), where: EY and £ 13 are the effective energy in MeV of the gamma and beta emission, ALiinh and ALiing are the most restrictive values of the annual limit of intake by inhalation and ingestion in Bq.
P. Rocco, M. Zucchetti /Journal of Nuclear Materials 233-237 ( 1996) 1500-1504
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Table l Clearance levels of some radionuclides relevant for fusion waste Radionuclide Halflife H-3 Be-10 C-14 Al-26 Si-32 Cl-36 K-40 Ca-41 Mn-53 Fe-55 Co-60 Ni-S9 Ni-63 Nb-93m Nb-94 Mo-93 Tc-99 Ag-108m Ag-llOm Sn-119m Sb-12Sm Ir-192M Ir-192
12.3S y l.6E6 y S730 y 7.16ES y 450y 3.0lES y 1.28E9 y l.4ES y 3.7E6 y 2.7 y 5.271 y 7.SE4 y 96 y 13.6 y 2.03E4 y 3.SE3 y 2.13E5 y 127 y 249.9 d 293 d 2.77 y 241 y 74.02d
As indicated previously, the analysis is performed after 50 y of interim storage for the most active components.
L(cf)• (Bq/kg) L(cc)b (Bq/kg) 3E6
2E6 4E4 2ES 4E2 lE4 4E4 SE3 7ES 6ES SES 4E2 3ES SES 2E5 6E2 9E4 IE4 6E2 4E2 SE4 2E3 6E3 IE3
5.1. In-vessel components
3ES
3ES
3E5 3E2 3E6 3E3
3E2
=clearance levels from a fining formula, see Eq. (1), taken from Ref. [9]. bL(cc) = clearance levels from a categorization scheme of Ref.
a L(cf)
[9].
Table 1 shows the clearance levels for radionuclides relevant for fusion waste, evaluated with the fitting formula and, when available, as a result of the categorization. It should be noted that: (!) Generally, the results of the categorization and of the fitting formula are similar. (2) With the exception of Al-26, Co-60, Ag-1 lOm, the clearance levels are more than 400 Bq/kg. The analysis performed in this paper on the declassification of waste is based on the evaluation of clearance indexes (or clearance ratings) Ic, such as: Ai
A;
A_
Li
L;
Lz,
The First Wall, Blanket and Back Plate have a total thickness of 48-56 cm in the inboard and outboard zone respectively. There are 0.2 cm of Be-coating and a 0.7 cm of Cu-Cr-Zr alloy in the FW, the other structural material is LN 316 SS. The surface dose rate of (outboard) Beryllium is 6 X 10- 4 Sv /h and the specific activity is 1 X 10 8 Bq/kg, 20% of which due to tritium. This result takes into account that the tritium inventory has been reduced by thermal treatments and that the residual activity of tritium, as in most materials, is less than 3.5 x 10 8 Bq/kg [15]. A further 16-fold reduction is due to 50 y of radioactive decay. Surface dose rate allows the remote handling recycling, provided that the main contaminant Co-60 (99.83% of the dose rate), deriving from 60 and 90 ppm of Ni and Co impurities can be removed. Beryllium in the inboard zone is impinged by a 10% higher flux, i.e., is almost in the same conditions and can be processed similarly. The surface dose rate of Cu-Cr-Zr is 2.7 Sv /h, which does not allow recycling. This material, as well that of the inboard zone ( 10% higher flux) should be disposed of. 316 LN SS of the outboard zone has surface dose rates varying from 1.6 to 1.2 X 10- 2 Sv /h. Substituting 316 LN SS with LA12TaLC (reduced activation ferritic steel) or V-5Ti, the maximum surface dose rates would be respectively 2.3 X 10- 3 and 4 X 10- 4 Sv /h [14] allowing recycling. The neutron flux impinging the inboard zone is from 4% to 85% higher. Also in this zone, reprocessing of alternative materials is feasible. Activation data for the diverter materials were not available, although waste management options should not differ substantially from those of the FW. 5.2. Vacuum vessel
I=-+ .. ·+-+ ... +__.::. c
where A; and L; are the specific activity and clearance level of the various radionuclides derived from the fitting formula of Ref. [9]. To allow the declassification to nonactive waste I 0 should be < I.
5. Application to ITER and extrapolation of results Activation and dimensional data for ITER materials are taken from Ref. [13], whereas extrapolated results for alternative materials are from Ref. [14].
The V.V. has 46-67 cm of thickness in the inboard and outboard zones respectively and consists of 3 cm-thick inner and outer layers of INCONEL containing a mixed array of 316 LN SS and INCONEL. Concerning 316 LN SS, the recycling of the alternative materials LA12TaLC is fully feasible. INCONEL in the outboard zone has a maximum surface dose rate of 4.16 X 10- 3 Sv /h. Almost 60% of the dose is due to Co-60 deriving from 900 ppm of Co impurity, 24% is due to Nb-94. arising from 3.56% of Nb, 16% from Co-60 arising from nickel of the alloy. Recycling should be feasible.
P. Rocco. M. Zucchetti /Journal of Nuclear Materials 233-237 ( 1996) 1500-1504
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INCONEL of the outboard outer layer has an activity of 5.5 X 10 4 Bq/kg, due to Ni-63 (91 %), Nb-93 (0.6%), Ni-59, (0.1 %), Nb-94 (0.05%) and minor contribution from other radionuclides. Ic evaluated as indicated previously is less than 0.7. This material could be declassified to non-active waste. The neutron flux impinging the inboard zone of the vacuum vessel is from 2.3 (inner layer) to 51 (outer layer) times higher than that of the outboard zone. All this material should be recycled, i.e. the outer layer could not be declassified. 5.3. Out-of-vessel components The outboard shield made with lead (3 cm) and 316 LN SS ( 1 cm) can be declassified to non-active waste. In particular, lead after one year of decay has an activity of 474 Bq/kg and Ic is less than 0.8 and could be declassified. The 316 LN SS layer after 50 y has an activity of 4500 Bq/kg and Ic = 0.9. The inboard shield is impinged by a 55 times higher flux. Lead could be declassified after less than 10 y, whereas the clearance index of 316 LN SS would be larger than 1. In order to allow declassification, this material should be substituted by reduced-activation steel. The outboard zone of the toroidal coil system is impinged by a neutron flux from 55 to 70 times lower than that on the inboard zone, so that the waste conditions are different. The outboard and inboard insulation (Cu, 0, C, Si) could be declassified after 1 and 50 y of decay, respectively. Similarly, the outboard Nb3Sn coils after 25 y of cooling have an activity of 3500 Bq/kg and Ic is less than 0.8, whereas this material in the inboard zone should be recycled. The outboard copper conductors can be declassified after 25 y (A = 688 Bq/kg and le = 0.24) whereas the inboard conductors should be recycled. The outboard casings in AISI-316 after 50 y of decay have A= 350 Bq/kg and le= 0.07. The declassification of the inboard casings could be attained with the use of reduced activation steel. Fig. 1 shows the different solutions described above.
6. Conclusions
The activation levels of materials used in ITER and values extrapolated to alternative materials have been used to analyze a strategy of waste management based on the recycling of the most active components and the declassification to non-active waste of the other ones. This analysis shows that both these options are feasible on alternative (i.e., reduced- or low-activation) materials or ITER-specific
Fig. I. Exploded view of an ITER sector, showing the various waste management options. (I) In-vessel components: Copper should be disposed of; Be and LAl2TaLC (in substitution of 316 LN SS) are recyclable. (2) Vacuum vessel: LA12TaLC is recyclable; INCONEL is either recyclable or can be declassified (outer layer of the outboard zone). (3) Out-of-vessel components: lead and insulation may be declassified; Nb 3 Sn coils may be recycled or declassified; outboard coil casings made with 316 LN SS may be declassified; inboard coil casings may be declassified if made with reduced-activation steel.
materials. These results should not change significantly if the materials investigated were subjected to the higher neutron fluencies of power fusion reactors. The purpose of this paper was to show that meaningful assessments on recycling and declassification of fusion materials can be performed from the analysis of materials used in an experimental fusion reactor like ITER. Further analyses on recycling-clearance should be carried out. In particular: (I) Concerning recycling, all information on the effective capability to separate by chemical-physical ways the noxious radionuclides from the bulk of the waste should be added to the dose rate related aspects. (2) The values of the clearance levels of significant radionuclides obtained with the fitting formula should be compared with the limits for near-surface disposal indicated in some national regulations, e.g., 1OCFR6 I [2]. It is also worth investigating a waste management scenario where the elimination of the noxious radioisotopes is the preliminary step for the declassification to non-active waste.
Acknowledgements
F. Toci was most helpful in the preparation of the paper. The figures were made by P. Bottelier and D. Caravati.
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