Activation Analyses of Disposal Options for Irradiated Be₁₂Ti

Abstract

The activity and disposal options for irradiated Be₁₂Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through an appropriate interface. The dedicated full-scale geometry MCNP model of the 11.25 degree HCPB DEMO torus was adapted to the requirements for the coupled 3D neutron transport and activation calculations. Special attention was paid to the use in the activation calculations of the commercial materials containing technological impurities. This has a crucial effect on the results and the impurities must be accounted for in any nuclear safety analyses. The short-term activity is formed by the radionuclides produced through the activation of Be and Ti nuclei and the long-term activity is formed by the products of the neutron irradiation taking place on the impurities. A prerequisite for the disposal or the recycling of the irradiated Be₁₂Ti is its deep detritiation; otherwise, the very high-tritium activity would fully prevent any attempt for its treatment. The most preferable is the use of the Be₁₂Ti with the composition including less material impurities, especially uranium. There could be the option to dispose the Be₁₂Ti intermediate-level wastes in the French repository after 1 year of cooling, assuming the detailed control of the impurities that fulfil the French authority requirements. The USA near-ground repositories could be an alternative to the European sites. The recycling of the irradiated Be₁₂Ti must first be elaborated and approved to ensure its treatment in a safe and efficient way. The remote handling technique must be developed for the re-fabrication of the Be₁₂Ti blocks.

1. Introduction

The development of a conceptual design for a demonstration fusion power plant (DEMO) has the highest priority in the EUROfusion Fusion Technology Department programme [1]. Two different concepts are now in progress to be implemented as a driver blanket for the upcoming DEMO fusion reactor: WCLL (Water Cooled Lithium Lead) [2] and HCPB (Helium Cooled Pebble Bed) [3]. Because of different reactor technologies and materials utilized, these two concepts have their own cons and pros affecting R&D activity. Pressurized water (155 bar) assumed in the cooling system of the WCLL DEMO even in spite of a complex chemistry seems to have some advantages compared to He gas coolant: the PWR-like technology is cheap, well established and robust. Liquid lead used as a neutron multiplier is widespread and not expensive. Weak points in the WCLL blanket concept are a corrosion between PbLi eutectic and steel structure, a tritium permeation from PbLi through steel walls in the water coolant and a very sophisticated tritium extraction technology. A solid breeder used in the HCPB blanket provides an efficient tritium generation and its extraction from the helium purge gas. The main drawback of the HCPB blanket technology is the very high capital expenses of this DEMO concept.

The cost factor is extremely important for public acceptance of the fusion facility and the fusion energy. The main disadvantages of the Be₁₂Ti neutron multiplier utilized in the HCPB breeder blanket are costs due to limited natural resources [4]. The price of raw beryllium increases continually with an annual rate of ~2.5%/year and it is estimated currently to be ~610 ÷ 650 USD/kg [4,5]. A conservative reasoning of the future price for the Be₁₂Ti industrial production for the DEMO needs could be estimated to be at least 1000 ÷ 2000 USD/kg in 30 years depending on the beryllium market evolution. Namely, the restricted Be resources are the reasons for the initiated research activity to find a reasonable solution for a possible Be recycling or, in the worst case, to define conditions for a possible disposal of the irradiated Be₁₂Ti. A total load of the Be₁₂Ti in the HCPB demo was estimated to be ~600 tons that corresponds to ~418 tons of Be [6]. In spite of the fact that the world’s identified resources of beryllium have been estimated to be more than 100,000 tons, the total world Be production does not exceed ~260 tons [5]. It is highly probable that the total Be amount required for the production of the Be₁₂Ti blocks for the HCPB breeder blankets will be provided from different global sources. Each Be ore possesses a specific chemical composition with various chemical element contents. As a result, the chemical composition of the Be (including impurities) used for the future HCPB breeder blankets cannot be defined exactly but it can be approximately assessed to make use of the available information for different Be ores.

An accumulation of radioactive nuclides in the Be₁₂Ti is inherent in the process accompanying its utilization in the HCPB DEMO blanket due to intensive neutron irradiation. Depending on the initial chemical composition of the Be₁₂Ti, its final inventory can differ significantly, affecting its treatment and processing strategy. This report presents an assessment of Be₁₂Ti activation in the current DEMO baseline model with the fully heterogeneous HCPB blanket concept reflecting the latest consolidated design and possible disposal options as well as outlines eventual recycling options.

2. Be₁₂Ti Neutron Multiplier

2.1. Beryllium Resources

Beryllium is currently produced from two minerals: bertrandite (Be₄Si2O₇(OH)₂) and beryl (Be₃Al₂Si₆O₁₈) [5]. Bertrandite, which contains about 15 wt% beryllium, is the principal beryllium mineral mined in the United States. Beryl, which can contain up to 5 weight percent beryllium, is the main beryllium mineral mined in the rest of the world. Beryl ore typically contains from 2 to 4 weight percent beryllium. Countries with active industrial beryl mining operations include Brazil, China, Madagascar, Mozambique, Nigeria, Portugal, Rwanda and Uganda [5].

The United States, Kazakhstan and China are the only countries known to process beryllium ores and concentrates into beryllium products [7]. Materion Corporation, Parkland Blvd., USA is the only producer of beryllium ore in the United States. The JSC Ulba Metallurgical Plant of Kazakhstan produces beryllium hydroxide from stockpiled concentrate previously extracted from the Ermakovskoe beryllium deposit in Russia [7]. The concentrate stockpile total accumulated by 1992; it was estimated that the Ulba Metallurgical Plant produced about 90 tons of beryllium in 2015 contained in beryllium–copper alloys, beryllium oxide ceramics, and beryllium metal. China’s Shuikoushan Non-Ferrous Metals Group Co., Ltd. mines beryllium ores to produce beryllium products in various forms at its refinery in Songbai, Hunan Province. The plant has an estimated production capacity of 150 metric tons per year of beryllium oxide (36 weight percent beryllium. The main beryllium world suppliers are listed in

Table 1. World mine Be production and reserves [5].

Country	Mine Production
USA	180
Brazil	3
China	70
Madagascar	1
Mozambique	13
Nigeria	1
Rwanda	1
Uganda	7
World total (rounded)	280

. In fact, the Materion (170 tons/year) and Ulba Plant (90 tons/year) are the major potential Be suppliers for the HCPB DEMO blankets.

The world annual Be production presented in Table 1 is significantly below the total needs for the future HCPB DEMO. Because of its chemical properties, beryllium metal is required in many industrial applications. A wide variety of products made from beryllium are used in aerospace, computer, defense, medical, nuclear and telecommunications applications. In the USA, a high-purity beryllium metal is classified as a strategic and critical material because of its vital importance for national security [7]. Therefore, it is very unlikely that the beryllium demands for the HCPB DEMO exceeding its annual world production can be covered without a long-term buying strategy. The appearance of such an important consumer as the DEMO on the global market could very likely be a driving force for the Be price increase. Hence the economical factor has to be accounted for when developing a strategy for the beryllium handling within the EUROfusion DEMO project.

2.2. Beryllium Chemical Composition

The impurities content in the original raw Be is the crucial issue for the future strategy of the industrial handling of the irradiated beryllium [6]. The concentration of various impurities significantly affects the results of the berillium activity accumulated during the irradiation phase. For the purposes of this work, the following assumption was made: the majority of the beryllium for the HCPB blanket manufacture will very likely come from the main suppliers, namely, from the Ulba (stands for JSC Ulba Metallurgical Plant of Kazakhstan) and/or from the Materion (which stands for Materion Corporation, Parkland Blvd., USA). The Be₁₂Ti will be produced utilizing beryllium: (1) from one of them, (2) separately from each of them or (3) from the mixture of them. Listed in

Table 2. Chemical compositions (wt%) of the Be from Ulba and Materion.

Element	Ulba	Materion
Be	balance	balance
BeO	-	0.9000
C	0.0500	0.0900
Fe	0.1000	0.0800
Al	0.0200	0.0500
Si	0.0300	0.0450
Mg	0.0200	0.0100
Zn	-	0.0050
Cr	0.0200	0.0100
Ni	0.0200	0.0250
Cu	0.0100	0.0250
Ti	-	0.0250
Zr	-	0.0050
U	0.0002	0.0150
Mn	0.0100	0.0050
Ag	-	0.0050
Co	-	0.0050
Pb	0.0500	0.0050
Ca	0.0050	0.0050
Mo	-	0.0050
O	0.8000	-
F	0.0010	-
Other	-	0.0400

are the chemical compositions of berillium from the Ulba, PTB-56 [8] and nuclear grade S-65 beryllium for the International Thermonuclear Experimental Reactor (ITER) from the Materion [9]. A chemical composition of Ti (RP-Ti grade OM-1, Polema [8]) is given in

Table 3. Chemical composition of Ti.

Element	Ti	Si	Fe	Ni	C	N	H
wt%	balance	0.10	0.20	0.20	0.05	0.08	0.35

.

An essential difference of the Be compositions presented in Table 2 is the uranium content. The presence of the uranium impurity is not of concern in case of the Be for the industrial applications without neutron irradiation. A quite different context is in nuclear facilities with the neutron irradiation where the utilization of beryllium will inherently result in an accumulation of α-radioactive and toxic transuranic isotopes of U, Pu, Am and Cm [10]. This will dramatically change the handling strategy of the irradiated Be₁₂Ti and its clearance according to IAEA recommendations [11] and could be unavoidable. Other impurities are also of high interest due to (a) the potential environmental risk and (b) the potential danger during a treatment.

2.3. Be₁₂Ti Chemical Composition

Eventually, Be for HCPB DEMO needs could originate not only from the main suppliers such as Materion and Ulba but rather from several other suppliers including minor providers. It is difficult nowadays to create an outline of the manufacturing strategy for Be₁₂Ti blocks. Generally, they can be produced either in different factories supplied by the domestic ores or Be could also be transported from various sources to collect and mix it in one place for its unified treatment and production. The latter option looks more feasible because it does not suppose a know-how transfer of the newly developed technology of the Be₁₂Ti blocks production [8]. Independently of the strategy, HCPB DEMO breeder blankets will be very likely filled with Be₁₂Ti blocks having slightly different material compositions. To account for the spread in the impurities content, three sets of the Be₁₂Ti chemical compositions (30%Ti and 70%Be in mass) were derived for the study: the composition with the minimum (Be₁₂Ti min) and maximum (Be₁₂Ti max) content of impurities and also the composition with the averaged content (Be₁₂Ti “average”),

Table 4. Chemical compositions of Be₁₂Ti with minimum, maximum and averaged impurities content, wt%.

Element	Be12Ti Min	Be12Ti Max	“Average”
Be	68.983 + 0.134	68.807 + 0.260	68.612 + 0.206
BeO	0.371	0.721	0.573
C	0.038	0.247	0.098
Fe	0.095	0.123	0.114
Al	0.005	0.035	0.019
Si	0.041	0.062	0.052
Mg	0.000	0.011	0.007
Zn	0.001	0.004	0.002
Cr	0.001	0.011	0.007
Ni	0.065	0.078	0.072
Cu	0.001	0.018	0.008
Ti	29.671	29.688	29.680
Zr	0.001	0.005	0.003
U	0.00014	0.0105	0.0021
Mn	0.000	0.005	0.003
Ag	0.000	0.004	0.002
Co	0.000	0.004	0.002
Pb	0.001	0.019	0.010
Ca	0.001	0.004	0.002
Mo	0.001	0.004	0.002
O	0.560 + 0.237	0.560 + 0.461	0.560 + 0.367
F	0.000	0.001	0.000
N	0.031	0.046	0.038
H	0.120	0.120	0.120
W	0.007	0.007	0.007
Na	0.001	0.001	0.001
Ta	0.001	0.001	0.001

. Therefore, the activation simulations with the irradiated Be₁₂Ti should be performed for these three material compositions to address the uncertainty with the impurities fraction.

2.4. Be₁₂Ti Blocks

The manufacturing method for the Be₁₂Ti blocks is based on the vacuum hot pressing of Be₁₂Ti powder [8]. Because beryllides are very hard materials, electrical discharge machining was employed to cut out flat surfaces and holes to form hexagonal blocks with a material density of 2.25 g/cm³ [8]. Shown in Figure 1 are the Be₁₂Ti blocks assembled in a unit according to the latest consolidated design of the HCPB blanket [3].

Radial cuts of the Be₁₂Ti unit shown in Figure 1 were proposed to implement in the HCPB blanket design to reduce thermal stresses that take place during the warming up and cooling down cycles. Such radial segmentation of the Be₁₂Ti unit can also be used for a mapping of the blocks in the blanket space to apply a practical classification of their properties after the thermal stresses and the neutron irradiation. This feature could be used in the future for the separate processing of the irradiated blocks to elaborate on a smart cycle with activated materials.

3. Modelling Methodology

3.1. HCPB Breeder Blanket Geometry Model

The HCPB blanket design based on the implementation of a Single Module Segmentation (SMS) conforms to the dimensions of the breeder blanket space in the generic DEMO CAD model for the DEMO baseline 2019 [12]. The blanket U-shaped first wall (FW) is 25 mm thick (20 mm in the HCPB blanket design 2017) and it is attached to a 150 mm thick back supporting wall (BSW), Figure 2. The blanket inner volume is subdivided into two parts separated by a back wall: a breeder zone (BZ) and a manifold. The manifold includes two big collectors for a He coolant and a small collector behind the back wall for a purge He gas. The BSW and gas collectors form the back supporting structure (BSS) of the HCPB blanket. The breeder units arranged in a hexagonal lattice, Figure 2, make up the breeder zone. The BZ is strengthened by radial central tubes of the breeder units ending with steel rods in the manifold region. The breeder ceramic annular layer, Figure 2, is surrounded by Be₁₂Ti blocks forming a hexagonal pyramid (see Figure 1). Breeder ceramic pebbles (Li₄SiO₄ + 35% mol. Li₂TiO₃, 60% ⁶Li enrichment) with 64% He volume fraction fill the cylindrical annual volume between the central feeding He pipe (∅ = 14 mm) and a separating pipe (∅ = 84 mm). A pressure (∅ = 94 outer, ∅ = 86 mm inner) and separating tubes form an annular gap for the outlet He coolant. The central and separating tubes have 2 mm thick walls. In the front part of the breeder ceramic layer, a Be₁₂Ti annular insert (∅ = 52 mm, 127 mm length) is arranged to enhance a neutron multiplication here, Figure 2. All structural elements of the HCPB blanket are made by EUROFER steel.

The MCNP geometry model of the HCPB blanket was developed on the module base: the breeder blanket volume of the generic model is filled with the first level universe representing the empty blanket housing or box (FW plus the back supporting wall); this universe in turn is filled with the second level one representing manifolds and empty BZ. The MCNP geometry model of the HCPB DEMO represents 11.25° toroidal sector including 1 inboard breeder (IB), 1 outboard lateral (OBL) and a half of the central outboard (OBC) blankets, Figure 3. To simulate a 360° DEMO tokamak reflecting boundary surfaces were implemented in the model. The total number of the MCNP geometry cells containing Be₁₂Ti is 425 and 410 in the IB and OBC blankets, respectively.

3.2. Computer Codes and Nuclear Data

The activation simulations are performed in two steps: (1) a particle transport simulation to assess the neutron spectra in the different locations, (2) inventory calculations based on the neutron spectra to quantify the generation of the radioactive nuclides.

Step 1 in this procedure requires neutron spectra in all geometry cells where further activation analyses will be carried out. To this end, the set of the geometry cells of interest must be specified in the MCNP input file to tally the neutron spectra. Special attention should be paid to the accuracy of the nuclear responses. This can be achieved by applying the proper variance reduction technique, if necessary. Neutron flux intensities and spectra required for the activation analysis were computed with MCNP6.2 [13], making use of the HCPB DEMO geometry model described above, the CCFE 709 neutron energy group structure [14] to address a fine resonance structure in nuclear data at low-neutron energies and the JEFF-3.3 nuclear cross-section data [15]. In the present study, only the activation of the Be₁₂Ti blocks was analyzed. Therefore, the tallying of the neutron spectra was performed in different geometry Be₁₂Ti cells that belong to the different poloidal segments of the blanket structure. The statistical uncertainty in all calculations of the total neutron flux does not exceed 1%. For some low-energy bins, for instance, 1⋅10⁻⁷~1⋅10⁻⁴ MeV, the relative error was calculated to be less ~10%. Normalization of the nuclear responses was carried out to the DEMO fusion nuclear power of 1998 MW, that results in a 7.094·10²⁰ s⁻¹ neutron source intensity.

To facilitate the activation analyses, an interface code was developed and applied to numerous activation calculations. It makes use of the MCNP input file for neutron transport calculations as a main repository to obtain full information on materials used in the calculations. The interface assures a proper preparation and an execution of the big amount of the FISPACT II [14] inventory calculations according to the geometry cell numbers specified in the MCNP neutron transport calculations. The FISPACT II code involves a TENDL activation data library which is TENDL-2021 in the present version [16]. For the numerous activation calculations, the interface automatically prepares the FISPACT input files for each geometry cell using a proper fine material composition and the irradiation scenario. It then automatically submits a batch job for each geometry cell for an execution. The whole procedure is fully parallelized making use of an MPI technique. The inventory results of the FISPACT outputs are collected and processed by a versatile post-processing code.

3.3. DEMO Operational Scenario

The current baseline DEMO design [1] assumes two phases of the reactor operation [17]. In the first phase (5.2 calendar years), the arrangement of a so-called driver blanket is assumed with a goal to achieve a maximum displacement damage of 20 dpa in the steel contained in the first wall of the breeder blanket. During the second phase (14.8 calendar years), the second breeder blanket will be used in the reactor tokamak where the steel structure can withstand at least 50 dpa. This study only considered the first phase, because the blankets currently under design are planned to be the driver ones for the DEMO. According to the irradiation schedule [17], the first phase only covers the first 5.2 calendar years of operation. Figure 4 shows a graphical representation of the irradiation schedule used in the activation calculations. The first DEMO irradiation phase assumes 5.2 years minus 10 days of DEMO operation at a 30% load of the nominal fusion power followed by the pulsed operation with 48 pulses of 4 h at the 100% power and 1 h dwell time in between. This results in 1.57 full-power years of the DEMO operation and the breeder blanket neutron irradiation exposure.

4. Results of Activation Calculations

4.1. Activity of the Be₁₂Ti in the HCPB Blankets

The activation calculations for the Be₁₂Ti blocks were performed in the IB and OB blanket modules of the HCPB DEMO to assess total activity and decay heat making use of the different Be₁₂Ti material definitions, Table 4. The DEMO tokamak is made up of 16 toroidal segments including 2 IB, 2 OBL and 1 OBC blankets. The total mass of the Be₁₂Ti in the HCPB DEMO, ~600 tons, was calculated making use of the MCNP geometry model and the results for different blankets are listed in

Table 5. Total mass of Be₁₂Ti in the HCPB DEMO blankets (tons).

IB Blanket	Central OB Blanket	Lateral OB Blanket	Total Reactor
6.8	8.3	7.9	603.2

. This corresponds to ~270 m³ of Be₁₂Ti in the HCPB DEMO tokamak.

During the first phase of the DEMO operation (5.2 CY), the total activity of the Be₁₂Ti accumulated in the HCPB OB blanket can achieve ~5.52·10¹⁷ Bq in the central outboard blanket (Be₁₂Ti max) right after the reactor shutdown, Figure 5. The spread of the results for the different material compositions, specified in Table 4, does not exceed ~1% for cooling times up to 1 year. The activity of the Be₁₂Ti drops down by a factor of ~50 one day and by the factor of ~500 one year after the shutdown of the reactor relative to the results right after the reactor shutdown. The difference of the results for the Be₁₂Ti (max) and (min) is about the factor of 6 and 2 at the moments of 10 and 100 years after the reactor shutdown, respectively. The results do not include tritium activity assuming its full detritiation after the shutdown. The uncertainty of the FISPACT inventory results was assessed only for the cooling phase and it amounts to 3% at the short cooling times (less than 1 year) and it is up to ~10% for the cooling times exceeding 1 year. The representative results of the Be₁₂Ti activity in the central outboard blanket (OBC) averaged over the Be₁₂Ti volume for the cooling times from the shutdown to 100 years after are given in

Table 6. Activity (Bq) of the Be₁₂Ti in the central OB blanket of the HCPB DEMO.

Be12Ti Composition	Cooling Time
	0	1 s	1 h	1 Day	1 Month	1 Year	10 Years	100 Years
Min	5.42 × 1017	2.54 × 1017	3.51 × 1016	2.70 × 1016	7.21 × 1016	1.07 × 1015	1.17 × 1013	5.02 × 1011
Average	5.44 × 1017	2.57 × 1017	3.62 × 1016	2.76 × 1016	7.42 × 1015	1.21 × 1015	4.51 × 1013	7.12 × 1011
Max	5.53 × 1017	2.65 × 1017	3.89 × 1016	2.91 × 1016	7.65 × 1015	1.35 × 1015	7.89 × 1013	1.17 × 1012

. The results discussed below in the text refer to the OBC blanket because the data for the lateral blanket are very close and the activity of the Be₁₂Ti in the IB blanket is ~1.8 times less compared to the results for the OB blankets.

The results for the decay heat power density in the Be₁₂Ti of the OB central blanket with different material specifications, Table 4, are presented in Figure 6. For the cooling time of up to 30 days, the decay heat power density for Be₁₂Ti with different impurities content differ insignificantly. After 100 years of cooling, the difference in the power density for the materials with minimum and maximum impurity concentrations reaches a factor of ~50. The power density right after the reactor shutdown is ~40 kW/m³ and it drops down by a factor ~100 30 days later.

The tritium production in Be₁₂Ti through ⁹Be(n,t)⁸Li reaction represents a potential danger in the handling procedure of any irradiated material including pure beryllium. Due to zero retention at high-operation temperatures [18], tritium is assumed to be removed from the Be₁₂Ti blocks with the He purge gas in the HCPB blanket [3] and therefore tritium activity is not accounted for in the present results (the results are nevertheless available). The tritium generation due to the fusion neutron irradiation of Ti is negligible and does not noticeably affect the activation of the Be₁₂Ti. The tritium retention in the Be₁₂Ti tends to be zero at temperatures above 600 °C and it goes up at lower temperatures [18]. The current HCPB blanket layout assumes that the Be₁₂Ti block is warmed up to ~800 °C in the regions close to the first wall and up to ~530 °C in the rear part of the breeder zone [3]. Due to an uncertainty of the experimental data [18], it is likely that some parts of the Be₁₂Ti assembly with not high enough temperatures could theoretically retain some tritium amount. This option is not considered in the current work.

The breakdown of the total activity of the Be₁₂Ti (“average”) in the central outboard blanket (OBC) is shown in Figure 7. Assuming that a remote handling of the “hot” blankets after the reactor shutdown might be possible after a specific cooling time, isotopes dominating the total activity at the cooling time longer than 1 month were determined and the breakdown of the total activity is presented in Figure 7. For the cooling time from 1 month to 1 year, ⁴⁵Ca (T_1/2 = 162.6 days) and ⁴⁶Sc (T_1/2 = 83.8 days) define the total activity. These isotopes are formed during the neutron irradiation of Ti due to (n,p) and (n,α) reactions. ⁶⁰Co (T_1/2 = 5.27 years) dominates the activity 10 years after the shutdown of the reactor; it is formed from the Co impurity in the Be₁₂Ti due to ⁵⁹Co(n,γ)⁶⁰Co reaction. The decay of ⁶³Ni defines the activity for the cooling time of 100 years. The significant concentration of ⁶³Ni(T_1/2 = 100 years) in the Be₁₂Ti results from the accumulation of this isotope due to the reaction on the Ni impurity, ⁶²Ni(n,γ)⁶³Ni.

A time evolution of the specific activity in the irradiated Be₁₂Ti of the different newly generated radioactive nuclides in the central OB blanket grouped according to atomic numbers is shown in Figure 8. The tritium removal bypass is switched off in the FISPACT activation simulation but tritium generation is not accounted for in the final result. Using the data presented in Figure 8, several groups of activated products responsible for the Be₁₂Ti activity can be identified according to their atomic numbers. Right after the reactor shutdown, a variety of radionuclides with a wide spread of atomic numbers are generated: light ones with A < 14, 18 < A < 30, 30 < A < 65 including fission products, 71 < A < 76, 80 < A < 82 and fissile nuclides with 90 < A< 97. One year after the shutdown, the main contributors to the activity are still nuclides with 18 < A < 30, some elements within the wide group of the fission products, Ta-W-Re, Tl and the group of the trans uranium elements: U, Pu and Am. After 100 years of decay, the most essential elements are N, Ni, Cd, U, Pu and Am. The chemical impurities available in the Be₁₂Ti affect and form its activity for all cooling times.

4.2. Disposal Options for the Be₁₂Ti

4.2.1. Total Disposal Limits

Each country with a current or future nuclear waste busyness develops different approaches to the handling of the nuclear waste: disposal, storage and recycling. Many countries are engaged in the elaboration of projects for the long-term disposal of radioactive wastes. The disposal facilities are developed to be operated on over many decades to keep their functionality up to hundreds of years. According to IAEA classification [19], radioactive waste is divided into very low (VLLV), low (LLW), intermediate (ILW) and high-level waste (HLW). The exact definition for the waste in each class differs from country to country. Two different methodologies for waste regulations are being applied to different repositories [20]: global activity limits to classify waste and limits on the accumulated individual radionuclides. A United Kingdom (UK) regulator employs global activity limits, France and Germany have adopted the limits for chosen isotopes, and Spanish and USA regulations assume limits for the nuclides additionally grouped in subclasses. The repository capacities in several countries for LLW (or near-surface disposal) waste are presented in

Table 7. The sizes of the repositories for LLW, [m³].

Country, Repository Name	Repository Capacity
United Kingdom, LLWR	1,800,000
France, L’Aube	1,000,000
Spain, El Cabril	36,000
Japan, Rokkasho	600,000
Finland, VLJ Oilkiluoto	8000
Sweden, SFR	266,000

[21]. Compared to the Be₁₂Ti volume in the HCPB blankets of ~270 m³, the available disposal capacities could be sufficient for its arrangement in a case of such decision.

The UK waste regulator applies global activity limits subdividing them into classes: maximum α-activity and β + γ activities with the upper limits of 4 MBq/kg and 12 MBq/kg, respectively [22,23]. The LLW criteria adopted in the Spanish regulations (low and intermediate level waste LILW limit level 1, El Cabril facility) amount to 0.185 MBq/kg and 37 MBq/kg for the α and β + γ activities, respectively [24,25,26,27].

Shown in Figure 9 is the specific β + γ activity of the Be₁₂Ti in the OBC blanket for three different material definitions (Table 3) compared to the available disposal limits. Respectively, the α-activity of the irradiated Be₁₂Ti is presented in Figure 10. The big spread of the results for the α-activity comes from the wide range of the uranium impurity concentration for different material compositions, Table 4. The data shown in Figure 9 and Figure 10 do not indicate the disposal opportunity for the irradiated neutron multiplier of the HCPB blankets in the repositories of UK and Spain due to the high β + γ activity. In case of an isotope separation procedure applied to the irradiated Be₁₂Ti, some disposal options could be considered for the α-radioactive nuclides.

4.2.2. Radionuclide-Specific Limits

Alternative to the total limits, the radionuclide-specific LLW limits operate with the data for separate radioactive isotopes that can be disposed in the dedicated radioactive waste repositories. The main difference between repositories is their depth. Two disposal opportunities are considered in this work: above ground (Centre de l’Aube in France, El Cabril in Spain, USA) and below ground (Konrad in Germany). The limits adopted in each country for the acceptance of the LLA disposal differ and they are defined by specific safety and technical requirements applied to the repositories.

Listed in

Table 8. Accumulated activity of various radioactive isotopes in the central OB blanket compared to the French nuclear waste regulations.

Radionuclide	LLW Limit (Bq/kg)	Be12ti Activity (bq/kg)
		1 Year	10 Years	100 Years
Sn-126	2.70 × 103	1.07 × 102	1.07 × 102	1.07 × 102
Cl-36	5.00 × 103	2.94 × 101	2.94 × 101	2.94 × 101
Nb-94	1.20 × 105	5.76 × 102	5.76 × 102	5.74 × 102
Th-232	1.10 × 106	6.50 × 10−6	6.50 × 10−6	6.50 × 10−6
Ag-108m	1.40 × 106	3.73 × 106	3.68 × 106	3.17 × 106
I-129	1.40 × 106	5.16 × 100	5.16 × 100	5.16 × 100
Be-10	5.10 × 106	4.81 × 105	4.81 × 105	4.81 × 105
Zr-93	1.80 × 107	8.99 × 102	8.99 × 102	8.99 × 102
Mo-93	3.80 × 107	1.17 × 105	1.13 × 105	1.11 × 105
Tc-99	4.40 × 107	1.13 × 104	1.13 × 104	1.04 × 104
Se-79	5.50 × 107	9.47 × 100	9.47 × 100	9.46 × 100
C-14	9.20 × 107	1.78 × 107	1.78 × 107	1.78 × 107
Ni-59	1.10 × 108	7.98 × 105	7.98 × 105	7.88 × 105
H-3	2.00 × 108	-	-	-
Cs-135	2.60 × 108	1.75 × 102	1.75 × 102	1.75 × 102
Ca-41	3.00 × 108	3.72 × 103	3.72 × 103	3.72 × 103
Pd-107	3.00 × 108	2.69 × 101	2.69 × 101	2.69 × 101
Cs-137	3.30 × 108	1.13 × 107	9.20 × 106	1.15 × 106
Sn-121m	3.70 × 108	1.62 × 104	1.40 × 104	3.39 × 103
Sm-151	4.50 × 108	3.39 × 105	3.16 × 105	1.58 × 105
Ni-63	3.20 × 109	1.06 × 108	9.93 × 107	5.34 × 107
Sr-90	6.00 × 109	5.30 × 106	4.27 × 106	4.89 × 105
Eu-154	5.80 × 1010	2.15 × 105	1.04 × 105	7.36 × 101
Eu-152	7.50 × 1010	5.24 × 103	3.30 × 103	3.28 × 101
Ce-144	8.80 × 1010	3.55 × 107	1.21 × 104	-
Ru-106	1.20 × 1011	3.78 × 107	8.36 × 104	-
Ag-110m	1.20 × 1011	7.37 × 108	8.15 × 104	-
Na-22	1.30 × 1011	5.79 × 106	4.36 × 105	5.78 × 10−5
Co-60	1.30 × 1011	1.28 × 1010	3.92 × 109	3.02 × 104
Cs-134	1.90 × 1011	2.70 × 106	1.31 × 105	1.58 × 10−7
Mn-54	3.60 × 1011	1.42 × 109	9.64 × 105	-
Sb-125	5.10 × 1011	3.31 × 106	3.46 × 105	1.38 × 10−4
Pm-147	5.80 × 1011	2.04 × 107	1.89 × 106	2.92 × 10−4
Zn-65	6.00 × 1011	1.95 × 108	1.75 × 104	-
Sn-119m	1.40 × 1012	6.77 × 103	2.86 × 100	-
Tl-204	3.30 × 1012	2.453 × 105	4.88 × 104	4.42 × 10−3
Fe-55	6.10 × 1012	1.28 × 1010	1.31 × 109	4.46 × 10−1

are the results for the accumulated specific activities for separated radionuclides in the irradiated Be₁₂Ti (“average” material composition) in the OBC blanket for different cooling times compared to the French maximum LLW acceptance limits for the Centre Stockage de l’Aube (CSA) disposal facility in France [22,28,29,30,31,32]. For the cooling time beyond 1 year, the activities of the accumulated isotopes do not exceed the LLW limits specified in Table 8 except ^108mAg (T_1/2=418 years). This isotope is produced as a result of the ¹⁰⁷Ag(n,γ)^108mAg reaction due to Ag impurity in the Be₁₂Ti and its activity exceeds the facility limit during the whole cooling time. In case of the Be₁₂Ti (min), material composition accumulation of this isotope is not expected, purification of the Be₁₂Ti (“average” and max) and extraction of the Ag impurity seems to be expensive and it is beyond the scope of this work. The fission products listed in Table 8 are not a safety issue, neither in the short term nor for the long-term waste disposal strategy. The volatile fission products such as ¹²⁹I also do not represent a risk for inhalation.

The Spanish national regulator also applies the radionuclide-specific criteria but it splits them into three subgroups: VLLW—very low-level waste and LILW—low–intermediate-level waste group 1 and 2. Presented in

Table 9. Accumulated activity of various radioactive isotopes in the central OB blanket compared to the Spanish nuclear waste regulations.

Radionuclide	LILW-1 (Bq/kg)	LILW-2 (Bq/kg)	Be12Ti Activity (Bq/kg)
			1 Year	10 Years	100 Years
Co-60	3.7 × 106	5.0 × 1010	1.28 × 1010	3.92 × 109	3.02 × 104
H-3	7.4 × 106	1.0 × 109	-	-	-
Cs-137	3.7 × 106	3.3 × 108	1.13 × 107	9.20 × 106	1.15 × 106
Nb-94	1.2 × 105	1.2 × 105	5.76 × 102	5.76 × 102	5.74 × 102
Ni-59	1.9 × 107	6.3 × 107	7.98 × 105	7.98 × 105	7.88 × 105
Ni-63	3.6 × 109	1.2 × 1010	1.06 × 108	9.93 × 107	5.34 × 107
C-14	6.1 × 107	2.0 × 108	1.78 × 107	1.78 × 107	1.78 × 107
Tc-99	3.0 × 105	1.0 × 106	1.13 × 104	1.13 × 104	1.04 × 104

are the waste acceptance limits for several isotopes adopted for the El Cabril disposal facility [25,26,32]. The LILW-1 level needs verification of a waste container resistance to compression and immersion with a certification for a possible diffused liquid release. The LILW-2 level must satisfy limits to leaching and thermal cycles. Both LILW levels assume homo- and heterogeneous waste incorporated or immobilized in a solid matrix. The specific activity of ⁶⁰Co and ¹³⁷Cs accumulated in the irradiated Be₁₂Ti could be an issue for the long-term storage in the El Cabril depository in case of the use of the container certified for the LILW-1 limit. In case of the Be12Timax composition, Table 2, the specific activities of these isotopes also do not exceed the El Cabril LILW-2 limits: 2.5⋅10¹⁰ and 5.7⋅10⁷ Bq/kg for ⁶⁰Co and ¹³⁷Cs, respectively. The total activity accumulated in the Be₁₂Ti of the HCPB breeder blankets 1 year after the shutdown compared to the installed capacity of the El Cabril facility [33] is presented in

Table 10. Total radionuclide accumulation in the Be₁₂Ti of the HCPB DEMO 1 year after the shutdown compared to the reference inventory (Bq) of the El Cabril disposal facility.

Radionuclide	El Cabril	Be12Ti
		IB	OBC	OBL	Tokamak
H-3	2.0 × 1014	-	-	-	-
C-14	2.0 × 1013	9.1 × 1010	1.5 × 1011	1.5 × 1011	1.0 × 1013
Ni-59	2.0 × 1014	4.2 × 109	6.6 × 109	6.5 × 109	4.3 × 1011
Ni-63	2.0 × 1015	5.6 × 1011	8.8 × 1011	8.7 × 1011	6.0 × 1013
Co-60	2.0 × 1018	7.0 × 1013	1.1 × 1014	1.1 × 1014	7.5 × 1015
Sr-90	2.0 × 1017	2.6 × 1010	4.4 × 1010	4.4 × 1010	2.9 × 1012
Nb-94	1.0 × 1012	2.5 × 106	4.8 × 106	4.8 × 106	3.1 × 108
Tc-99	3.2 × 1012	5.9 × 107	9.4 × 107	9.3 × 107	6.4 × 109
I-129	1.5 × 1011	2.5 × 104	4.3 × 104	4.3 × 104	2.9 × 106
Cs-137	3.7 × 1015	5.5 × 1010	9.4 × 1010	9.4 × 1010	6.3 × 1012

. The total activity of ¹⁴C accumulated in the HCPB tokamak is about a half of the total capacity of the El Cabril facility for this radionuclide and it seems to be too high for the disposal there. The entire or partial disposal in the El Cabril depository of the irradiated Be₁₂Ti could be nevertheless feasible.

Within a KONEKT research project [34], repository options for the irradiated beryllium were analyzed to assess possibilities of its storage in Germany. A total mass of the irradiated Be already disposed in different institutions and locations in Germany is estimated to be ~3000 kg [34]. The assessment of the safety issues of the Be₁₂Ti activation results produced in this report is performed making use of the facility limits available for the Konrad mine repository. Although this repository is big enough to arrange millions of tons of radioactive materials, its capacity for the storage of the radioactive Be is limited to 24.5 kg (Table 12 [35]) because of the potential danger of groundwater contamination. A theoretical option for the Be₁₂Ti secure disposal in the steel container of Type I [35] with a useful volume of 3.9 m³ was chosen for the comparison assuming the limits for the waste group APG03 (metal products, Table 5 and Table 6 [35]). Presented in

Table 11. Accumulated activity of various radioactive isotopes in the central OB blanket compared to the German nuclear waste regulations for the Konrad repository.

Isotope	Limit, Bq/m3	Activity, Bq/m3	Isotope	Limit, Bq/m3	Activity, Bq/m3
Be-10	1.0 × 1012	1.1 × 109	H-3	6.9 × 1014	-
Cl-36	1.0 × 1012	6.7 × 104	C-14	8.7 × 1012	4.1 × 1010
Ar-39	2.1 × 1012	8.2 × 108	Na-22	2.8 × 1012	1.1 × 1010
Ca-41	6.4 × 1011	8.5 × 106	S-35	9.7 × 1013	3.7 × 105
Fe-55	1.1 × 1015	2.9 × 1013	Ca-45	1.6 × 1014	1.8 × 1014
Co-60	2.0 × 1012	2.9 × 1013	Sc-46	7.9 × 1012	6.8 × 1013
Ni-63	2.8 × 1013	2.4 × 1011	V-49	2.2 × 1015	7.0 × 1010
Rb-87	1.4 × 1012	3.8 × 100	Cr-51	8.2 × 1014	4.3 × 108
Nb-94	1.9 × 1011	1.3 × 106	Mn-54	1.2 × 1013	3.2 × 1012
Ag-108m	1.0 × 1012	8.5 × 109	Co-57	1.2 × 1013	1.3 × 1013
Sn-126	1.3 × 1012	2.5 × 105	Co-58	1.8 × 1013	1.2 × 1012
Cs-137	3.3 × 1012	2.6 × 1010	Fe-59	1.4 × 1013	1.8 × 109
Pb-210	5.6 × 1011	4.8 × 101	Ni-59	3.8 × 1013	1.8 × 109
Ra-226	1.8 × 1010	3.9 × 100	Zn-65	1.8 × 1013	4.5 × 1011
Ac-227	9.7 × 1010	3.0 × 101	Se-79	6.7 × 1012	2.2 × 104
Ra-228	1.4 × 1011	1.7 × 10−1	Kr-85	1.6 × 1013	1.3 × 109
Th-230	7.4 × 109	2.5 × 103	Sr-89	3.6 × 1013	1.0 × 109
Pa-231	7.7 × 109	3.4 × 102	Sr-90	2.6 × 1012	1.2 × 1010
Th-232	5.1 × 109	1.5 × 10−2	Nb-93m	1.3 × 1014	1.1 × 108
U-233	6.9 × 109	2.5 × 103	Mo-93	3.1 × 1013	2.6 × 108
U-234	1.0 × 1010	6.0 × 105	Zr-93	4.1 × 1012	2.1 × 106
U-235	5.6 × 109	2.5 × 104	Nb-95	3.1 × 1013	2.3 × 1010
Np-237	1.3 × 1010	1.2 × 106	Zr-95	2.2 × 1013	1.1 × 1010
U-238	2.0 × 1010	5.6 × 105	Tc-99	3.1 × 1012	2.6 × 107
Pu-238	3.3 × 1011	1.9 × 109	Ru-103	4.4 × 1013	7.2 × 108
Am-242m	1.4 × 1010	2.1 × 106	Ru-106	5.4 × 1012	8.6 × 1010
Pu-244	1.8 × 1010	4.7 × 10−5	Pd-107	2.0 × 1013	6.2 × 104
Cm-245	3.3 × 1010	2.7 × 101	Cd-109	7.9 × 1013	1.0 × 1011
Cm-247	1.4 × 1010	2.2 × 10−7	Ag-110m	3.8 × 1012	1.7 × 1012
Cm-248	1.2 × 1010	8.1 × 10−8	Cd-113m	1.3 × 1013	3.7 × 106
			I-125	1.1 × 1014	4.0 × 102
			Sb-125	9.7 × 1012	7.6 × 109
			Te-125m	1.4 × 1014	1.8 × 109
			I-129	2.6 × 1012	1.2 × 104
			Ba-133	7.9 × 1012	2.9 × 105
			Cs-134	4.1 × 1012	6.2 × 109
			Cs-135	3.6 × 1012	4.0 × 105
			Ce-144	7.4 × 1012	8.1 × 1010

are the results for the activity of the accumulated isotopes in the irradiated Be₁₂Ti (“average”) 1 year after the reactor shutdown compared to the Konrad facility limits. The available results for tritium accumulation are not included in Table 11 because the full tritium removal from the irradiated Be₁₂Ti blocks is foreseen and expected in the HCPB blankets as it is pointed out above. The accumulated activities of ⁶⁰Co (produced from Co impurity) and ⁴⁶Sc (produced from the Ti irradiation) exceed the Konrad limits by the factor of ~10. The results for ⁴⁵Ca and ⁵⁷Co slightly exceed the limits. In case of several years of decay for ⁴⁶Sc, ⁴⁵Ca and ⁵⁷Co isotopes, the safety limits for their disposal can be reliably secured. In the case of the ⁶⁰Co storage in the Konrad repository, the decay time of several decades is necessary to meet the safety requirements.

The classification of waste for near-surface disposal in the USA supposes three classes: A, B and C [36]. The waste of class A must satisfy the minimum safety requirements and it includes mildly contaminated waste products (trash, clothing, etc.). The waste of classes B and C can be applied to nuclear reactor facilities, exhausted radioactive sources, etc. Class B must meet rigorous requirements to ensure stability after disposal. Class C assumes not only rigorous requirements and provides stability but also it must assure protection against inadvertent intrusion [36]. The specific activity of several long-lived radioactive isotopes (T_1/2> 5 years) in the irradiated Be₁₂Ti of the OBC blanket after 1 year of the cooling time is displayed in

Table 12. Specific activity (Bq/m³) of several long-lived radionuclides in the Be₁₂Ti compared to the limits for the different classes of LLW in the USA [36].

Radionuclide	Class A	Class C	Be12Ti
C-14 in activated metal	2.96 × 1011	Class A × 10	4.1 × 1010
Ni-59 in activated metal	8.14 × 1011	Class A × 10	1.8 × 109
Nb-94 in activated metal	7.40 × 108	Class A × 10	1.3 × 106
Tc-99	1.11 × 1011	Class A × 10	2.6 × 107
I-129	2.96 × 108	Class A × 10	1.2 × 104
Pu-241	1.30 × 104 1	Class A × 10	2.2 × 104 1
Cm-242	7.40 × 104 1	Class A × 10	1.7 × 102 1

¹ Units are Bq/g.

compared to the limits for different classes of the LLW in the USA. Class B is not used in this specification. The results for the short-lived isotopes (T_1/2< 5 years) 1 year after the shutdown are presented in

Table 13. Specific activity (Bq/m³) of several short-lived radionuclides in the Be₁₂Ti compared to the limits for the different classes of LLW in the USA.

Radionuclide	Class A	Class B	Class C	Be12Ti
H-3	1.5 × 1012	(1)	(1)	-
Co-60	2.6 × 1013	(1)	(1)	2.9 × 1013
Ni-63 in activated metal	1.3 × 1012	2.6 × 1013	2.6 × 1014	2.4 × 1011
Sr-90	1.5 × 109	5.6 × 1012	2.6 × 1014	1.2 × 1010
Cs-137	3.7 × 1010	1.6 × 1012	1.7 × 1014	2.6 × 1010

¹ There are no limits established for this radionuclide in Class B or C wastes. This waste shall be Class B unless the concentrations of other nuclides in this determine the waste to Class C independent of these nuclides.

. The disposal of the irradiated Be₁₂Ti could likely be feasible in the USA near-surface facilities. The concentrations of ²⁴¹Pu and ⁶⁰Co are not a hindrance for the storage of the irradiated Be₁₂Ti because the limits and corresponding measures for class B or C should then be applied for the containers.

4.2.3. Location Dependent Processing of the Irradiated Be₁₂Ti Blocks

The prismatic Be₁₂Ti unit has a radial length of ~42 cm in the outboard blankets in the reactor mid plane, Figure 1. Because of the block structure of this unit, there could be an opportunity to process the differently located Be₁₂Ti blocks under the strict remote control. This enables the special treatment of the irradiated blocks located in the different radial positions in the Be₁₂Ti assembly making use of a remote handling technique. Figure 11 displays the results for the β+γ-specific activity of the Be₁₂Ti assembly (“average” material composition) arranged in the OBC blanket for two radial segments: a front zone (close to FW) of ~26 cm length and a rear zone of ~16 cm (close to back wall). The results averaged over the total volume of the Be₁₂Ti in the OBC blanket (“average”, Figure 9) are also presented for the comparison. One year after the reactor shutdown, the specific activity for the front zone is higher by ~35% and that for the rear zone is lower by ~70% compared to the “average” results. Due to the FW neutron wall-loading profile, the upper, central and bottom parts of the blanket are exposed to the different neutron fluxes; in the mid plane of the reactor, the neutron flux is at a maximum. To assess the difference of the results for these parts, the specific activity was calculated in the highest, central and lowest locations in the front zone of the Be₁₂Ti assemblies (26 cm length). The β+γ-specific activity in the “front” part of the Be₁₂Ti assemblies 1 year after the reactor shutdown in the upper and bottom parts of the blanket are lower compared to the central part by ~15% and ~23%, respectively. The results indicate a significant difference for the specific activity of the Be₁₂Ti along the radial distance from the FW. The poloidal distribution of the activity along the blanket height is not so large. In case of the disposal possibility, the irradiated Be₁₂Ti blocks could be processed separately using a “layer by layer” approach to find the best processing strategy.

4.3. Dominant Radionuclides

The detailed breakdown of the generated activity and dose rates due to inhalation for different cooling times is given in

Table 14. Dominant radionuclide in the Be₁₂Ti of the OBC blanket responsible for activity and dose rates (inhalation) for different cooling times.

Isotope	Activity, Bq/kg	Isotope	Dose Rate Inhalation, %
1 s
He-6	2.52 × 1013	Sc-46	3.80 × 101
Sc-48	2.22 × 1012	Sc-48	2.23 × 101
Sc-47	7.54 × 1011	Ca-45	1.25 × 101
Sc-46	6.12 × 1011	Sc-47	5.04 × 100
Sc-46m	4.82 × 1011	Co-60	4.14 × 100
Ca-45	3.71 × 1011	Pu-239	2.19 × 100
N-16	2.03 × 1011	Ta-182	1.72 × 100
Ag-110	1.31 × 1011	He-6	1.68 × 100
		Xe-133	1.53 × 100
		Sn-130	1.03 × 100
		Sn-129m	8.55 × 10−1
		Pu-238	8.39 × 10−1
		Tc-105	7.01 × 10−1
		Sn-129	6.81 × 10−1
		Pu-241	4.77 × 10−1
1 h
Sc-48	2.18 × 1012	Sc-46	4.12 × 101
Sc-47	7.48 × 1011	Sc-48	2.38 × 101
Sc-46	6.11 × 1011	Ca-45	1.36 × 101
Ca-45	3.71 × 1011	Sc-47	5.41 × 100
W-187	7.58 × 1010	Co-60	4.49 × 100
Sc-49	5.14 × 1010	Pu-239	2.37 × 100
Np-239	3.71 × 1010	Ta-182	1.86 × 100
Mn-56	3.51 × 1010	Xe-133	1.66 × 100
Ti-45	3.42 × 1010	Pu-238	9.09 × 10−1
Co-58m	2.03 × 1010	Pu-241	5.17 × 10−1
		Xe-135	4.88 × 10−1
1 day
Sc-48	1.51 × 1012	Sc-46	4.52 × 101
Sc-47	6.17 × 1011	Sc-48	1.83 × 101
Sc-46	6.07 × 1011	Ca-45	1.50 × 101
Ca-45	3.69 × 1011	Co-60	4.96 × 100
W-187	3.88 × 1010	Sc-47	4.93 × 100
Np-239	2.80 × 1010	Pu-239	2.62 × 100
Ta-182	1.87 × 1010	Ta-182	2.05 × 100
Co-58	1.78 × 1010	Xe-133	1.81 × 100
Fe-55	1.65 × 1010	Pu-238	1.01 × 100
		Pu-241	5.72 × 10−1
30 days
Sc-46	4.77 × 1011	Sc-46	5.75 × 101
Ca-45	3.26 × 1011	Ca-45	2.14 × 101
Fe-55	1.62 × 1010	Co-60	7.94 × 100
Ta-182	1.57 × 1010	Pu-239	4.26 × 100
Co-60	1.45 × 1010	Ta-182	2.78 × 100
Co-57	1.34 × 1010	Pu-238	1.64 × 100
Co-58	1.34 × 1010	Pu-241	9.21 × 10−1
W-185	3.32 × 109	Pu-240	5.77 × 10−1
		Co-58	4.98 × 10−1
1 year
Ca-45	7.85 × 1010	Co-60	2.81 × 101
Sc-46	2.99 × 1010	Ca-45	2.06 × 101
Fe-55	1.28 × 1010	Pu-239	1.70 × 101
Co-60	1.28 × 1010	Sc-46	1.44 × 101
Co-57	5.70 × 109	Pu-238	6.52 × 100
Ta-182	2.07 × 109	Pu-241	3.52 × 100
Mn-54	1.42 × 109	Pu-240	2.30 × 100
Ag-110m	7.37 × 108	Kr-85	1.55 × 100
		Ta-182	1.46 × 100
		Ar-39	9.64 × 10−1
		Fe-55	7.00 × 10−1
		Ag-110m	6.26 × 10−1
		Am-241	5.58 × 10−1
10 years
Co-60	3.92 × 109	Pu-239	4.15 × 101
Fe-55	1.31 × 109	Co-60	2.10 × 101
Ni-63	9.93 × 107	Pu-238	1.48 × 101
C-14	1.78 × 107	Pu-240	5.62 × 100
		Pu-241	5.56 × 100
		Am-241	5.50 × 100
		Ar-39	2.30 × 100
		Kr-85	2.12 × 100
		Ar-42	9.99 × 10−1
		Fe-55	1.75 × 10−1
100 years
Ni-63	5.34 × 107	Pu-239	6.09 × 101
C-14	1.76 × 107	Am-241	1.70 × 101
Ag-108m	3.17 × 106	Pu-238	1.07 × 101
Pu-239	2.00 × 106	Pu-240	8.19 × 100
U-235m	2.00 × 106	Ar-39	2.68 × 100
Cs-137	1.15 × 106	Ar-42	2.22 × 10−1
Ba-137m	1.09 × 106
Ni-59	7.88 × 105
Am-241	6.96 × 105
Y-90	4.89 × 105
Sr-90	4.89 × 105
Be-10	4.81 × 105

. The results were obtained with Be₁₂Ti “average” material composition in the central OB blanket and they do not include the data for tritium accumulation due to its removal from the blanket with the purge gas. The threshold for the inclusion of the isotopes in the table is ~0.5% from the total activity or dose rate. The short-term activity of the Be₁₂Ti is dominated by Sc and Ca isotopes accumulated through (n,xpxn) reactions on Ti isotopes. Therefore, the inherent presence of these isotopes in the radioactive wastes has to be tolerated and the waste handling should account for this issue. The long-term activity of the Be₁₂Ti is formed mainly by the nuclear reaction products coming from the neutron irradiation of the impurities available in the initial material. To some extent, this issue could be managed through (a) the purification of the Be and Ti used for the Be₁₂Ti production and (b) through the use of more “pure” materials such as Be₁₂Ti from the Ulba factory, Table 2. At the very long-cooling storage time (>100 years), the fissile isotopes and fission products also appear in the group of the dominating isotopes.

5. Recycling Possibilities for the Irradiated Be₁₂Ti

For the DEMO fusion reactor, the two operation phases are considered with durations of 5.2 and 14.8 calendar years. Because of the limited world Be resources, a recycling option should also be considered to relax or decrease the Be demands for the HCPB DEMO blankets in the second phase. Any recycling strategy has to include the following three generic steps: (1) detritiation, (2) purification or removal of the radioactive products and (3) production of the Be₁₂Ti blocks.

The basic procedure for the tritium removal from the irradiated Be₁₂Ti blocks is high-temperature annealing. The experiments performed at the irradiated Be samples show that tritium release has a non-diffusive (explosive) nature [10]: the tritium release reaction rate as a function of the annealing temperature has a sharp peak. The effective temperature of the peak depends on the neutron fluence applied to the samples: the higher the fluence, the more the tritium atoms accumulate and the lower the peak of the annealing temperature. The sample heating rate also affects the peak temperature: the increase in the rate leads to the decrease in the peak temperature [10]. We expect that the detritiation of the Be₁₂Ti blocks follows the same tendencies. The temperature of the peaks for an efficient tritium release varies from ~400 to 900 °C depending on the irradiation history and the heating rate. The peak value of the tritium concentration in the irradiated Be₁₂Ti calculated in the equatorial mid plane of the blanket was assessed without implementation of the tritium removal rate and it was estimated to be ~1.9⋅10¹² Bq/g right after the reactor shutdown. This value drastically exceeds the clearance limit of 100 Bq/g [11]. It exceeds by ~10,000 times and ~2000 times the limits accepted in the Centre Stockage de l’Aube and in the El Cabril facilities, respectively (Table 8 and Table 9). Therefore, the deep and long detritiation procedure seems to be absolutely necessary to assure the safe treatment of the irradiated Be₁₂Ti. The duration of the annealing procedure to reach the acceptable level of the tritium concentration is beyond the subject of this paper.

In the case of the full recycling of the irradiated Be₁₂Ti, the purification process should follow the detritiation of the blocks. Nowadays, the technology for the recycling of the irradiated Be and, moreover, for the irradiated Be₁₂Ti, does not exist in an industrial scale. The attempts for purification of the normal metallic Be were already performed making use of different technologies: vacuum distillation, zone melting and chlorine processes [37]. The chlorine method includes the chlorination of beryllium and then the reduction of the beryllium chloride to pure metallic Be. Beryllium halides are fairly volatile at high temperatures (up to 1500 °C) and they are much more volatile compared to the impurities in halides. This route is used to obtain high-purity beryllium. The degree of the purification depends on a number of cycles applied: dissolution, precipitation and washing. The decontamination factor could vary from ~10² ÷ 10⁴ and be higher depending on the cycle number (1 ÷ 3) and the duration of the whole procedure could take up to 1 year [38]. The final product very likely will still contain some residual radioactivity preventing its conventional treatment without clearance. This refining process is very energy consuming and additionally it requires a significant amount of the dangerous acid. Overall, this procedure is harmful and in general, environmentally unfriendly. To avoid the additional environmental burden, a special purification factory should be constructed close to the DEMO site. This factory should be naturally protected against any radioactive leakage to the environment. It should also be fully automated and the remote handling has to be widely implemented. Applied to the irradiated Be₁₂Ti, the purification process seems to be very expensive resulting in a significant price increase for the final product. Whether the purification of the irradiated Be₁₂Ti is worth developing for the DEMO materials’ recycling is beyond the scope of this work.

The last step in the theoretical recycling route of the irradiated Be₁₂Ti is a refabrication of the blocks. If the purified Be₁₂Ti is available as in the result of the previous step, the manufacturing of the new blocks follows the already well-established technology from the fresh Be and Ti. After the purification process, remote handling would be necessary to exclude the personnel radiation exposure. Because of the residual activity, the remote handling for the manufacture of the new blocks and blankets afterwards looks to be unavoidable. The purification step could be fully skipped. In this case, after the detritiation, the irradiated Be₁₂Ti blocks would be directed to the refabricating facility. The secure and well-protected facility against operational and environmental contamination seems to be conceivable or even theoretically feasible. Such a plant has to include a pressing machine for the crushing of the Be₁₂Ti blocks and another one for the refabricating of the blocks. The remote control and handling are therefore necessary for the final blanket fabrication. This theoretical option could only be feasible if the irradiated Be₁₂Ti is allowed to be used for the HCPB blanket fabrication and the proper technology is developed, approved and certified for the secure application.

6. Conclusions

The activity and the disposal options of the irradiated Be₁₂Ti were assessed for the HCPB DEMO blankets making use of a code system that enables performing 3D activation calculations by linking the Monte Carlo transport code MCNP and the fusion inventory code FISPACT through the appropriate interface. The dedicated full-scale geometry MCNP model of the 11.25 degree HCPB DEMO torus was adapted to the requirements for the coupled 3D neutron transport and activation calculations. Special attention was paid to the use in the activation calculations of the commercial materials containing technological impurities. This has a crucial effect on the results and the impurities must be carefully accounted for in any nuclear safety analyses.

The activation analyses were performed for the HCPB DEMO breeder blankets for several decay cooling times from shutdown up to 100 years. The method used enabled the calculations for the activity of the Be₁₂Ti in different locations in the HCPB blankets. The total activity of the Be₁₂Ti (“average” material composition) in the whole reactor is ~5.42·10¹⁷ Bq right after the reactor shutdown and it decreases up to ~7.2·10¹⁵ Bq after 1 month. For the longer cooling time, the differences in the total activity of the Be₁₂Ti with various material compositions become more significant and the activity has the deviation for different material compositions. The short-term activity is formed by the radionuclides produced through the activation of Be and Ti nuclei and the long-term activity is formed by the products of the neutron irradiation taking place on the impurities.

A prerequisite for the disposal or the recycling of the irradiated Be₁₂Ti is its deep detritiation; otherwise, the very high-tritium activity would fully prevent any attempt for its treatment. The most preferable is the use of the Be₁₂Ti from the Ulba factory: this material composition includes fewer material impurities, especially uranium. The uranium impurity should be kept at the level below 0.002 wt.% to satisfy the safety limits for the storage. The disposal of the Be₁₂Ti blocks irradiated in the HCPB DEMO is not possible 100 years after the reactor shutdown in the UK repositories. There could be the option to dispose the Be₁₂Ti intermediate-level wastes in the French repository after 1 year of cooling assuming the detailed control of the impurities that fulfil the French authority requirements. The fission products resulting from the uranium impurity do not represent the environmental risk, for both short- and long-term disposal strategies of the irradiated Be₁₂Ti. The El Cabril repository could also be the proper site for the disposal of the irradiated Be₁₂Ti if the integral limits will not be applied in this case. The German repository even in spite of its large capacity does not provide enough space for the Be₁₂Ti wastes’ storage. In the case of the theoretical possibility for the Be₁₂Ti disposal, the safety limits applied in Germany prevent its storage in the repositories. The USA near-ground repositories could be an alternative to the European sites. The tritium retention in the irradiated Be₁₂Ti has to be well investigated and assessed to control the tritium activity and to develop the disposal strategy to ensure secure disposal options. The Be₁₂Ti blocks irradiated in the HCPB DEMO blankets could be securely disposed in several repositories worldwide (for instance, in France or in the USA) in case of the use of the Be ore with low-uranium impurity (for example, Ulba beryllium resources).

The well-elaborated and certified recycling option of the irradiated Be₁₂Ti blocks is not yet addressed and developed. The remote handling technology for the re-fabrication of the blocks from the irradiated material has to be well tested, approved and certified. The location for the manufacturing facility should be close to the reactor site to avoid the transportation of the activated materials over well-populated regions. Such a procedure will significantly increase the capital DEMO costs and probably result in a public rejection of the whole project. In light of these heavy conditions, the recycling of the activated Be₁₂Ti seems to be questionable and it requires significant investigations of its practical feasibility.