First Wall Components

Chapter 7

First Wall Components Igor V. Mazul, Georgij L. Saksagansky JSC D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, Saint Petersburg, Russia

Chapter Outline 7.1 Introduction 211 7.2 First-Wall Design Principles 216 7.2.1 Design Algorithm 216 7.2.2 Initial Stage Design 219 7.2.3 Estimation of the Engineering and Physical Characteristics of the First-Wall Components 220 7.3 ITER First Wall 226 7.3.1 First-Wall Components 226

7.3.2 Component Modelling: Technological and Testing Facilities 7.3.3 Prevention of Destructive Events 7.4 Next-Generation Reactor First Wall 7.4.1 Challenges 7.4.2 Possible Engineering and Physical Solutions 7.5 Alternative Uses of First-Wall Technologies References

230 234 236 236 237 245 245

7.1 INTRODUCTION The first wall (FW) of a fusion reactor includes a set of in-chamber components, or structures, located within the chamber cavity and facing the plasma. This term also refers to the interior surface of the reactor closest to the plasma. Shaping the plasma profile, the FW acts as a shield for the magnetic fusion reactor (MFR) systems, protecting them from particle and radiant energy fluxes emitted by the plasma, excluding neutrons. The FW structural materials and design solutions must meet the MFR integrated optimisation design criterion for minimising the generation and introduction to the plasma of extrinsic (‘non-hydrogen’) particles. A reflection of how important it is for the fusion science and plasma physics community is the fact that issues of concern

Fundamentals of Magnetic Thermonuclear Reactor Design. http://dx.doi.org/10.1016/B978-0-08-102470-6.00007-X Copyright © 2018 Elsevier Ltd. All rights reserved.

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related to this criterion have been known as the ‘first-wall problem’ since as early as the 1960s. During a stationary discharge phase, high-density plasma occupies the core of a plasma column. Because the plasma confinement by magnetic fields is not ideal, H ions and α-particles drift from the core towards the wall, cutting across the magnetic field lines, losing energy and getting recharged (through charge-exchange reactions) on their way. Hitting the wall, they produce a double harmful effect of, first, thinning (by sputtering) the wall and, second, cooling the plasma by radiation losses through poisoning the plasma by wall erosion products. To use the expensive superconducting magnetic system (that accounts for 20%–25% of the total tokamak capital cost) to the best advantage, the edge plasma layer between the plasma column surface and the FW is made only 3–4 cm thick in large operating tokamaks and up to 10 cm in the international thermonuclear experimental reactor (ITER). This ring-shaped layer is held in place by the limiters—FW structural components located along the toroidal circumference of the chamber. The particle and heat flows hitting the FW are extremely severe (see Table 7.1 [1, 2]), especially near the limiters and divertor targets. Disruption of the discharge current (when the instantaneous values of heat flow density may reach 10 GW/m2) is a big hazard. One important point is that a simple understanding of the physical processes taking place in the plasma does not necessarily mean that you can provide a detailed prediction of the plasma characteristics needed to design new, larger-scale machines. Existing scalings have an empirical base and cannot ensure the quality of ‘long-shot’ extrapolations. One should also bear in mind that in ITER, the plasma column size is several-fold, and its volume is an order of magnitude greater than in the precursor JET tokamak. The prediction uncertainty is particularly relevant to cases where a prediction of current disruptions, Edge Localized Mode (ELM) effects (instabilities developing at the plasma edge) and runaway electron effects is to be made, and where the number of prospective events, as well as heat load magnitudes, lengths and locations are to be modelled. Thus, data concerning the FW components’ operating environment should be treated as tentative, and this assumption should be given precedence in the design of the FW components. To perform the protective function, the FW must not be transparent to ­p article and heat flows coming from the plasma. In other words, it must have optical integrity and use a cooling system. In this respect, the graphite liner s­ urrounding the plasma column and re-emitting heat coming from the plasma can hardly be regarded as an absolute solution: while confining the plasma spatially and shielding nearby structures from incident particle fluxes, the liner offers no protection against the heat flows.

TABLE 7.1 Expected Operating Conditions of the ITER In-Chamber Components (EDA Phase, 2001) Operation conditions

First wall

Start-up limiter

Divertor vertical target

Baffle (the upper part of the divertor vertical target)

Divertor liner

1

2

3

4

5

6

  Quasi-stationary (MW/m2)

0.5

0.5

10

1–3(5)

<1

  Transient (MW/m2)

—

4–8

20

  Vertical displacement (MJ/m )

60

—

—

60

—

  Plasma current disruption (MJ/m2)

0.36

1

12

1

1–3

—

—

25

—

—

Peak heat load

2

2

  ‘Runaway’ electrons (MJ/m )

—

  A quasi-stationary load

400/3000

400/3000

400/3000

400/3000

400/3000

  Transient conditions

—

30/6000

10/300

1/3000

—

  Vertical displacement

0.2/150

—

—

0.2/150

—

  Plasma current disruption

10−3/3000

10−3/3000

10−3/300

10−3/300

10−3/300

10−2/300

  Under the runaway electron effect

—

—

170

15

100–136

∼20

∼10

Component area (m )

680

7

50

150

50

Radiation damage expected level (dpa)

1.6 (Be) 5.3 (Cu) 2.7 (SS)

1.6 (Be) 5.3 (Cu) 2.7 (SS)

<1(Cu)

<1(Cu)

<1(Cu)

Incident heat flow total power (MW) 2

213

(Continued)

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Pulse length(s)/number of heat pulses during

214

Operation conditions

First wall

Start-up limiter

Divertor vertical target

Baffle (the upper part of the divertor vertical target)

Divertor liner

1

2

3

4

5

6

1016–1017

1017–1018

1018–1020

1017–1018

1017–1018

  Particle flow energy (eV)

∼100

∼100

∼100

∼100

∼10

Expected surface erosion rate (mm/year1)

< 0.5

0.5

∼5

<1

<0.1

Projected hours of operation till replacement/tolerable number of replacements

4600/-

2300/2

∼500/3–8

∼500/3–8

∼500/3–8

Immediate contact with plasma/ limited position error (m)

Short term/±2

Yes/±0.5

Yes/± (1–3)

Short term/±1

No/±5

  Surface (MW/m2)

0.78

0.78

3 (Cu)

5 (Cu)

3 (Cu)

  Volumetric heat (MW/m3)

4.6 (Be) 7.3 (Cu)

4.6 (Be) 7.3 (Cu)

Be

Be

CFCa

W

W

Quasi-stationary particleflux parameters Ion/c-x neutral flux density (cm−2 s−1)

Neutron load

Armour material a

Carbon-fibre composite.

Fundamentals of Magnetic Thermonuclear Reactor Design

TABLE 7.1 Expected Operating Conditions of the ITER In-Chamber Components (EDA Phase, 2001) (cont.)

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The said integrated criterion entails a number of meaningful requirements for the FW materials and structures. These include ‘inertness’ with respect to plasma and the small inventory/capacitance with respect to fuel. Each of the said properties refers to a wide range of physical and chemical phenomena, and therefore, maybe a brief comment would be worthwhile. The FW ‘inertness’ refers to the FW’s physical and technological ability to introduce as little as possible impurity species into the plasma. Impurities in fusion plasmas have detrimental effects, the most important being the increase in energy losses and the reduction of produced fusion power (Section 6.2). To ensure a desired inertness to FW, it is necessary to establish the physical, technological and temperature conditions allowing a decrease in ion- and heat-induced erosion and use wall coating/armour containing low-Z elements, such as Li, Be, B and C. In this respect, the use of tungsten (W) for ITER divertor targets is questionable. While being sputtering resistant, W has a high atomic number, which makes the ultimate effect of its use ambiguous. The parameter used to evaluate the inertness is ρi Zi2 (where ρi is the coefficient of sputtering of an armour with a Zi atomic number by plasma ions/atoms), which has to be minimised. The FW fuel ‘capacitance’ problem is a difficult one and has not been solved satisfactorily yet. The tricky aspect is the H isotope accumulation in the FW nearsurface layer as a result of sorption and implantation of incident ions. This problem is particularly acute with respect to the radiotoxic and very expensive tritium. In short-cycle experimental devices, the sorption and diffusion processes may directly affect plasma’s behaviour. The sorption of fuel onto the ‘clean’ wall depletes the plasma during a discharge, while desorption and an abrupt H ejection from a gas-saturated wall may lead to a loss of stability. In long-cycle reactors, two processes take place at a time: (1) the fuel mix external ejection and (2) uncontrolled sorption onto a continuously regenerated getter film, deposited on the wall through limiter sputtering. These two processes determine the fuel mix current balance. Tritium accumulation in wall materials, meaning a withdrawal from the fuel cycle of more than 1 kg of tritium—a pretty large amount by fusion technology standards—is an issue of critical importance to the MFR. This factor is detrimental to the reactor’s technical and economic metrics and increases the operational risk. It is this risk that constrains the use of the graphite armour, despite the material’s low atomic number, heat resistivity and technological effectiveness. The reason is the very high sorption activity of carbon films freshly deposited on the walls during reactor operation. One possible solution to this problem is the operation mode optimisation and transition to new materials (Chapter 13). Although the technological achievements of the past few decades are indisputable, the first-wall problem is still there. The weaknesses that remain are the limiters and the targets’ short durability and the need for their replacement, tritium accumulation in the nearsurface and the re-deposited layers, and the high cost of the FW components. Therefore, the evolution of technological and physical solutions is an inevitable

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upward process for the fusion industry. One cannot, for example, rule out the revision of current concepts of energy absorption in favour of liquid-metal or movable divertors (Section 7.4). It is not only the first wall of a fusion machine that has to sustain severe heat loads combined with other destructive impacts. Key functional components of jet and rocket engines, electrical-ionisation lasers, nuclear reactors, magnetohydrodynamic (MHD) generators, powerful electrical vacuum devices, plasma torches, and heat-to-electricity conversion plants are exposed to similar stresses. For this reason, the closing subsection of this chapter addresses the alternative uses of the first-wall technology.

7.2  FIRST-WALL DESIGN PRINCIPLES 7.2.1  Design Algorithm In the course of their 60 years of evolution, the first-wall components were getting more diverse and acquiring new functions. In the 1st-Gen tokamaks, plasma is only separated from adjacent systems by the vacuum (discharge) chamber wall, and there are stationary ring-shaped diaphragms inside the chamber (Fig. 7.1). The diaphragms, following the shape of the plasma column, set the confined plasma boundary. Ions emitted from that area and travelling along

FIGURE 7.1  1st-Gen tokamak structural scheme. 1 – vacuum (discharge) chamber; 2 – edge plasma area; 3 – ring-shaped diaphragms; 4 – plasma column.

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the toroid helically are bound to hit the diaphragms and neutralise, and their lifetime is much shorter than that of the plasma ions. The diaphragms also protect the walls against ‘runaway’ electrons. The wall reacts and absorbs all other energy flows. Because of the low intensity of those flows and the small thermal deformations, the 1st-Gen tokamak FW and vacuum boundary performed the same functions. With plasma discharge getting more powerful and longer, the idea was, first, to curtain off the wall by heat-accumulating elements, and then to chill it down using forced water cooling. The divertor configuration was launched in the 1980s. We remember that the divertor’s basic operating principle is to deform magnetic field lines in such a way as to force the direct contact between the edge plasma and the wall to occur as far from the plasma column as possible (Figs 7.2 and 7.3). In a divertor design, edge plasma field lines stop having poloidal symmetry. Instead, they are tilted towards the divertor chamber, making up a separatrix, that is, a boundary magnetic surface separating the plasma column from the edge plasma. The separatrix crosses the energy-receiving divertor, dumping out the larger part of charged particles and heat flows on it. A well-regulated drainage system enables a substantial decrease in the intensity of particle and heat flows hitting the discharge chamber walls (in ITER the decrease is approximately 2×). In addition, the edge plasma, where secondary particles get ionised and return to the receiving target along the divertor field lines, shields the plasma column.

FIGURE 7.2  Tokamak fragmentary sectional view featuring the divertor (ITER EDA phase, 2001).

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FIGURE 7.3  Divertor configuration of ITER reactor (CDA phase, beginning of 1990s) [3]. q(x) is the density of heat flow hitting the divertor target.

Therefore, the divertor reduces impurity content, prohibiting impurities from the wall from entering the confined plasma. Another important purpose of the divertor is to take helium ash and impurities out of the reactor by concentrating them closer to cryosorption pumps and intensifying their removal. The divertor configuration is an example of how the design of a sophisticated electrical physics device can be optimised to give te most effective division of technological functions between different elements. The previous chapter addressed this issue in relation to the vacuum system. Here we shall touch upon some thermal physical aspects. The walls of the discharge and divertor chambers together make up a non-uniformly loaded toroidal structure located inside other reactor structures (the cryo-

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stat, the toroidal and poloidal field coils, the shields, etc.), which are partially or completely closed. It is impracticable to withdraw it for maintenance or replace as a whole structure, but it is possible to repair local damage of the armour or replace several panels with the help of a robot. This aim may be achieved by drastically decreasing the heat load on the discharge chamber’s first wall and increasing the heat load on sacrificial (replaceable) divertor targets. Let us explain why. The divertor targets receive around 50% of the α-particles’ power, which significantly reduces the density of the heat flux acting on the discharge chamber walls. In addition, the targets receive charged particles responsible for sputtering, slowing up the ion erosion. Finally, there are reasons to believe that the divertor protects the walls, at least partially, against heat loads caused by runaway electrons and plasma current disruptions. The heat and particle fluxes diverted from the discharge chamber walls are directed towards the divertor targets. The targets’ surfaces, being much smaller than the walls’ surfaces, experience many times greater specific loads. There are still no technical or technological solutions to divertor targets that can withstand such loads throughout the reactor lifetime. Therefore, the ITER design provides for a remote replacement of the targets. The targets’ small size and location in the lower part of the reactor facilitate their handling. The ratio between the discharge-wall-bound and divertor-targets-bound heat fluxes can be varied using simple procedures, such as an automated programmable fuel and impurities mix introduction into the edge plasma. The variation interval is 20% to 80%.

7.2.2  Initial Stage Design The design of the first wall or any other complex engineering system calls for a stage-by-stage approach to solving the optimisation problems. At the initial stage, one has to take conceptual decisions, while the methods for their engineering implementation, as well as relevant parameters and characteristics are developed/defined at later stages. The first step to be made at the initial stage is to determine the amount and the most likely distribution of the heat fluxes coming from the plasma to the FW components. In tokamaks, this amount is the sum total of ohmic and additional heating powers. In ITER, its peak value is determined by the α-particles and additional heating, while in DEMO, the main constituent is the α-particles power, accounting for ∼20% of the fusion power. Different FW components are exposed to peak loads at different times during the operation cycle. The divertor target and the discharge chamber walls are subjected to the highest loads during the stationary discharge stage, while for the limiter the loads peak during the discharge initiation and shut-down. The initial design stage is mostly concerned with quasi-stationary processes that are longer than 5–10 s and strongly affect the FW’s composition. The effects produced by disruptions and other pulse and transient processes are a­ ccounted

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for at later stages. The second step to be made at this stage is to better define the FW layout, arrangement and acceptable size. It is desirable to increase the FW surface to decrease the specific heat load. However, this desire contradicts the pursuit of the best possible technical and economic characteristics of the reactor as a whole. The space extending radially from the plasma core near the equatorial plane is particularly insufficient, and it may be feasible to place the energy receiving part above and/or under the plasma column. Placing the targets further away from the X-point may help reduce the heat loads and slow down the ion sputtering. The seemingly best solution to remove the targets out of the magnet system is unviable due to the divertor’s poloidal configuration. Probably, this could be achieved with a bundle divertor. However, the use of the latter would give rise to new intricate problems. Issues related to the replacement of the FW components should also be considered in the FW’s design. For example, the bulky limiter panels in the equatorial region can reasonably be placed outside the plasma column where the widest vacuum ports are located. The third step is to assess the consistency of the selected FW option with the goal of achieving the highest possible economic performance and energy effectiveness. A reactor’s economic performance is measured by its availability factor, depending on the duration of outages needed for the replacement of in-chamber components (ICCs) in accordance with the schedule. The experimental ITER machine provides for such an opportunity. However, a demonstration reactor design does not necessarily allow periodic and even one-off replacements of operating parts to be made, and the divertor concept is likely to be substantially different from that for the ITER. The availability factor also affects the susceptibility to radiation damage, determining the range of candidate structural and functional materials. The energy conversion efficiency depends mostly on the coolant. This is why high-temperature coolants, including helium and liquid metals, are preferred. At the end of the initial design stage we determine the following: l l

ICCs’ relative positions and limit sizes maximum loads to which ICCs are exposed under quasi-stationary operating conditions l frequency and method of ICCs replacement l most suitable coolant l candidate materials and their neutron irradiation doses

7.2.3  Estimation of the Engineering and Physical Characteristics of the First-Wall Components The purpose of the second design stage is to define the FW’s compositional structure. To this end, one has to l l

estimate heat loads on different components, define coolant’s parameters,

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l l

select the material for the heat-sink panel and armour, compute the temperature conditions and determine the armour’s highest acceptable thickness and erosion lifetime, and l estimate the strength and functional fatigue of the structures.

7.2.3.1  Heat Load Estimation A stationary surface heat load on the first wall includes plasma’s electromagnetic radiation and charge-exchange atoms. Its total value is usually taken to be 20% of the fusion power (excluding fusion neutrons power). The average specific heat load (total power load on the wall divided by the wall area) is generally taken to be at least 2× greater, considering the heat flows’ potential spatial non-uniformity. The vertical divertor targets and the limiter face the plasma directly. Plasma heat is transferred to the wall along magnetic field lines, and the specific load depends on the longitudinal (along a field line) heat flux density and the target tilt angle. Heat distribution across the field lines is determined for physical reasons. In a simple case, including a low-recycling fusion experiment, when energy losses from the edge plasma are low, this distribution can be set analytically. In the case of a gaseous divertor target, when the inert gas density near the target is high, computational codes can be used for analysis purposes. The changing poloidal angle between the wall and a field line is the main tool for controlling the heat flux density, as shown in Fig. 7.3. The smaller the angle, the less is the density (and the greater the space needed for the target), but practically, the range of these variations is limited. Both the target with flat areas and gaps between the cassettes and a rippled magnetic field are nonuniform. At a poloidal angle less than 10–15 degrees a peaking heat load density increase due to this factor may surpass the benefit of the average density reduction. 7.2.3.2  Determination of Coolant’s Parameters With a given coolant type and a given heat load, there is some ambiguity about the selection of parameters for the optimum coolant as different factors come into play. For the sake of distinctness let us consider ITER, which uses water as a coolant. We start with estimating the minimum acceptable inlet water temperature. This provides headroom for the temperature acceptable for different parts of the structure, and thereby increases the structure’s functional longevity. For water, the minimum temperature could be 30°C, but for ITER, it has been raised to be 100°C. The reason is that at lower temperatures, the main candidate materials for the cooling panels (steel and copper alloys) experience radiationinduced brittleness—even at small radiation doses. A high inlet temperature helps avoid this risk. It is also for this reason that the inlet temperature of a coolant (helium or a liquid metal) is increased to 600°C–700°C, if the heatremoving panels are made of molybdenum or tungsten.

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The following factor should be considered in the selection of a coolant’s pressure and speed. When using water or another two-phase (liquid–vapour) coolant, measures should be taken to avoid a general thermal crisis that may lead to a thermal destruction of the structure. In other words, the channel’s wall temperature, Tw, must be lower than a critical temperature, Tcrit, which depends on the coolant pressure. Therefore, the pressure should be selected such as to preclude any near-wall boiling or at least make it controllable. To this end, the jet cooling technique and devices based on the vapotron effect are employed. Wall (coolant side) temperature is defined by the equation Tw = Tc + q/α where Tc is the coolant’s temperature, q is the heat flow density and α is the heat transfer coefficient. This is the only parameter that can be controlled by completely physical means and one that is the greater the better. It increases with the speed of the coolant, and also with the flow turbulence degree and the channel’s inside area. The most straightforward way to increase α is to speed up the coolant flow. The limitations of this method are the increase in pressure within the cooling system, the greater erosion of the pipes and the greater consumption of energy needed to make the flow run. The erosion becomes a critical factor at a coolant flow speed higher than 10 m/s. For these reasons, the best way to intensify the cooling process is to add some sort of fins or spikes, for example, on the inner surface of the channels to increase the surface area in contact with the coolant and enhance the near-surface microturbulence.

7.2.3.3  Material Selection The problem of structural and functional materials is addressed in Chapter 13. We only note here that for ITER structures using forced cooling, the materials of choice are the SS316 austenitic stainless steel (if heat loads are less than 1 MW/m2) and the CuCrZr alloy (if the loads are greater). 7.2.3.4  Estimation of the First-Wall Thickness and Temperature Field As the FW layers have different functions, their thickness criteria are different as well. The thickness of a load-bearing element is determined by strength requirements. It has practically no effect on other layers in terms of their thermal physical ‘condition.’ The load-bearing element is only subject to relatively small heat loads caused by fusion neutrons. For this reason, its cooling channels are incorporated in the ‘tail’ of the cooling circuit of the heat sink panels. The largest heat loads on those panels come from the armour side. It is therefore important to minimise the interfacial joint temperature. As the temperature of the plasmafacing surface is limited by the armour’s material properties, one can make the armour thicker to improve its erosion lifetime. The interfacial joint temperature is affected not only by the coolant temperature, but also by the thickness and shape of the cooling channel segments closest to the plasma. The main require-

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ments for the cooling channels are strength and leak-tightness. We remember that in ITER, the coolant’s pressure limit is 4 MPa in normal conditions and 6 MPa in test conditions. Channels with ∼2-mm walls seem to be suitable. For a flat FW with round-cross-section cooling channels, the maximum joint temperature is determined by the interchannel space and distance from a channel centre to the joint point equidistant from neighbouring channels. These distances must be as small as possible. To achieve this, it would be good to have cooling channels with a rectangular cross section. So, minimum joint temperature Tmin, depending on heat load, is in the 100°C–400°C window for different FW components. The material and thickness of the armour tiles are selected using the following criteria. Surface maximum temperature, Тmax, is limited by the flux of particles it emits, with evaporation not necessarily being a critical factor. For example, for graphite armour, which is prone to self-sputtering and chemical erosion, the maximum surface temperature must be within ∼1500°С (although it could well be increased to 2000°C if the mechanical strength considerations prevailed). If Be and W tiles are used, Тmax can be estimated based on requirements such as resistivity to thermal erosion and recrystallisation leading to loss of strength, as well as some technological factors. A series of thermal tests have proved that the acceptable temperature is 1000°С for Be tiles and 2500°С for W tiles. After selecting a Tmax and ignoring the negligibly small temperature difference in the interfacial joint layer, we determine the greatest admissible tile thickness: h1 ~ (Tmax - Tmin ) × λ (T ) / q where λ(T) is the tile’s average thermal conductivity in the Tmax – Tmin range. Table 7.2 summarises the results of a thermal–physical modelling of the ITER components subject to the most severe heat fluxes, that is, the start-up limiter (q = 8 MW/m2) and the divertor’s vertical target (q = 20 MW/m2).

TABLE 7.2 Characteristics of Different Armour Materials Characteristic

Material Be

CFC

W

Max allowable surface temperature (Tmax (°С))

1000

1500

2500

Average thermal conductivity in the Tmax–Tmin range (W/m·K)

110

180

110

Min temperature of armour contacting with a heat sink panel (°C) q = 8 MW/m2

300

300

300

q = 20 MW/m2

450

450

450

9.6

27

30

3

9.45

9.6

Max allowable tile thickness (mm) q = 8 MW/m2 2

q = 20 MW/m

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7.2.3.5  Armour Erosion Lifetime The longevity of ITER’s most heat loaded in-chamber components is limited by the armour erosion. Tiles grow thinner in the course of reactor operation and finally become unable to protect the heat sink panel. When calculating the thinning rate, the following parameters should be accounted for: l thermal erosion due to surface overheating l sputtering l in-chamber transfer of eroded particles

during transient events

The physical mechanisms of thermal erosion are quite diverse. The list includes, first, evaporation. The total amount of evaporated matter depends on the surface temperature and duration of the surface heated state, as well as on the pressure of saturated vapours in the ‘working’ temperature range. The heat load baseline values in Table 7.1 are not enough to estimate the heating kinetics of evaporating surfaces. In processes involving fast energy exchange, (e.g. a current disruption) a plasma cloud of evaporated material appearing near the wall (with a pressure of up to several MPa) helps shield the surface. The result is a many-fold (10× and more!) reduction of the heat load on the wall and a lower evaporation rate. It is difficult to analyse this phenomenon qualitatively, and experimental data are needed to verify relevant computational simulations. Second, heat acting on graphite and similar materials may give rise to macroscopic erosion, when grains and even pieces are emitted from the material. Mass lost due to macroerosion is estimated at 10–100 µm per each current disruption event. As several hundred current disruptions are projected to occur during ITER’s operation, the evaporated layer may become thinner by tens of millimetres. Third, the surface melting and the melted layer removal by electromagnetic forces, the plasma ‘wind’ and the MHD-instabilities in a liquid-metal film may contribute much to thermal erosion. The fourth physical mechanism of erosion, ion sputtering, may be regarded as the most significant. For example, a graphite armour layer lost due to sputtering during ITER operation is estimated to be up to 10 m (!?) thick. However, these estimates ignore one notable factor. Atoms leaving the wall surface as a result of sputtering get ionised and are deposited back to the wall (the redeposition process). The ion sputtering rate can be estimated realistically with sophisticated code packages for a 3D kinetic modelling of sputtered particle transport, such as the REDEP [5]. The use of these better precision tools results in 10× to 100× lower sputtering rate in regions subject to the highest loads. Thus, the maximum theoretical value of a sputtered off graphite wall layer can be conservatively put at 102–103 mm. With erosion being the prevailing factor determining the FW’s lifetime, the tiles should be as thick as practically feasible, the only constraint being the material’s temperature limit.

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7.2.3.6  Strength and Fatigue Lifetime The stress–strain state of the FW panels is the result of a number of loads, the most important being the following: l l l l

thermal mechanical stresses electrodynamic loads baric stresses caused, for example, by coolant’s pressure process loads

Thermal mechanical loads occur in compound structures whose parts have different thermal expansion coefficients, and also can be generated by nonuniform temperature fields. These stresses are proportional to the elastic (Young’s) modulus and a material’s thermal expansion coefficient and are inversely proportional to its thermal conductivity. Therefore, structures using forced cooling systems should have thin walls, because, with increasing thickness of the wall, load-bearing capacity grows slower than the thermal mechanical stresses, while the surface receiving the heat flow gets hotter. In addition, a material’s strength and thermal conductivity decrease with increasing temperature. The remaining constraint is the coolant’s pressure of 3–4 MPa. Generally, coolant channel wall is in the range of 1–5 mm. A free deflection of the wall towards the heat flow could reduce the thermal mechanical stresses. However, this option is unacceptable for the FW components for two reasons: the panels’ load-bearing element must be sufficiently bulky to resist electrodynamic loads and sufficiently strong to prevent any uncontrollable deformation of components arranged around the toroidal circumference. This is important to ensure a uniform distribution of heat flows over the wall surface. An important point about the optimisation design of multilayer fragments is that contacting materials have close thermal expansion properties. An example is the Be-CuCrZr pair. Where it is impossible, dissimilar materials are separated by a high-plasticity intermediate layer, such as pure copper. This is how tungsten-armoured FW components are made. The copper intermediate layer absorbs the stresses that are developed at the border with tungsten and reduces them drastically at the junction with the heat sink panel. The electrodynamic loads are due to interaction between eddy currents occurring in the FW elements and tokamak magnetic fields. They can be simulated using, for example, the TYPHOON code package (Chapter 4). We know that electromagnetic forces depend on the rate of transient magnetic events, structure face layer’s electric conductivity and the contour size. For ITER, the maximum electromagnetic force pressure is within ∼4 MPa. One way to decrease them is to make structures electrically segmented by making full thickness incisions. Residual process stresses occur in multilayer elements subjected to mechanical deformation, such as bending, or to cooling down after a high-temperature operation. Gravity loads on the FW components are negligibly small.

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The FW stress–strain state is generally analysed using 3D simulations (Chapter 12). Lifetime projections are necessary for structural components containing forced-flow cooling channels. Such components should have an at least 2× margin for projected deformations and a number of operation cycles should be at least 20× smaller than that derived from the low-cycle fatigue curve. During the final design stage, a detailed design work and optimisation aimed at a better definition of system design specifications and the ICC engineering and physical characteristics should be carried out. This includes the consideration of issues related to safety analysis (e.g. minimisation of tritium accumulating in the armour/coolant and particulates deposition in slots), as well as the issues of fabrication, installation, mounting and dismantling of different components, prevention of the loss of coolant, destruction and loss of the armour tiles and other emergencies.

7.3  ITER FIRST WALL Numerical simulation, experimental adjustment and comprehensive testing of the ITER’s first-wall components were part of an extensive international research collaboration programme [4]. This work will continue during ITER’s operation. The scope of the said programme is governed by the complexity, uniqueness and high cost of the FW components, and the fact that it is impossible to reproduce experimentally at the ‘pre-reactor’ stage or to simulate the destructive factors comprehensively. The hallmarks of this research are the following: l l

Its highly interdisciplinary, integrative nature. The technical solutions’ variability and high probability of implementation due to the many prospective modes of the reactor operation and FW components replacement while the reactor is still operating. l Utilisation of large-scale electro-physical devices and special equipment intended for modelling or simulation of the plasma’s negative effects on the first wall; and opportunity to use the research site as venue for in-chamber component manufacturing operations.

7.3.1  First-Wall Components The ITER divertor uses 54 water-cooled ‘cassette assemblies’, each of which has a supporting structure that carries plasma-facing components, that is, the inner and outer divertor targets, the liner and the dome (Fig. 7.4). The cassette assemblies act as a common water collector for PFCs and local neutron shield for the vacuum chamber. The divertor design is optimised such that three requirements are met: (1) no direct contact between a cassette body and plasma, (2) minimal flow into the plasma of the neutral gas generated due to divertor plasma neutralisation and (3) a vacuum duct to pumps with sufficient gas-kinetic conductivity. A divertor target has two distinct parts: the upper, bent one (the baffle), and the lower, straight one, crossed by the separatrix (the vertical target).

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FIGURE 7.4  Parts of the ITER divertor (EDA phase, 2001).

The vertical targets are in contact with the plasma throughout the discharge stationary phase absorbing up to 80% of the α-particle flux. The dome and the baffle are directly exposed to plasma for only 10 s, while the divertor configuration is shaped up. At this stage, the separatrix keeps sliding on their surfaces. Only the plasma’s electromagnetic radiation and the re-ionisation neutrals hit the dome and the baffle during the stationary phase. The liner is not exposed to the divertor plasma at all, and the intensity of radiant energy and particle fluxes hitting it is several times lower compared to what the dome and the baffle are exposed to. With the divertor operating effectively throughout the stationary phase of the operation cycle, the only function left for the ITER limiters is to protect the discharge chamber’s first wall during the discharge start-up and shutdown. For this reason, they are often referred to as start-up limiters (Fig. 7.5). The start-up phase begins from a gas breakdown. After that, a toroidal cold plasma column develops around the outer circumference of the discharge chamber. As the longitudinal plasma current increases under the combined action of the plasma poloidal magnetic field and the field due to the poloidal coils, the divertor

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Fundamentals of Magnetic Thermonuclear Reactor Design

FIGURE 7.5  The ITER start-up limiter and first wall [4].

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configuration shapes up. This takes around 60 s. All this time, the plasma is in direct contact with the limiters. As there are only two limiters in the reactor, it is important to ensure a uniform heat flux distribution on the limiters’ surfaces. To this end, the surfaces have to be carefully formed. The same considerations are valid when imposing very tight requirements on the limiter fabrication precision and positioning accuracy. The peak heat load on the limiters is up to ∼8 MW/m−2 during the start-up and the final stages of the operation cycle. At other stages, the thermal conditions for the limiters and the FW are more or less the same. The discharge chamber area as measured around the plasma column is close to 544 m2 (∼80% of total projection area); the divertor chamber baffle and the port limiter occupy around 77 and 9 m2, or ∼11% and 1.3%, respectively. The rest of the near-plasma space is occupied by ports allowing additional heating and diagnostics of the plasma, as well as a remote maintenance of in-chamber components using robotic technologies. One a priori effective design approach to ensuring ITER availability, radiation safety and ecological sustainability is to rely, wherever possible, on well-tested and robust concepts, techniques, processes and materials. For unique experimental projects, to which ITER belongs, following the engineering conservatism principle is an imperative. It is for this reason that the ITER design adopted the traditional tokamak first wall configuration: the static solid structure with a leaktight forced-cooling circuit, consisting of three layers and enabling a remote replacement—whether scheduled or emergency-driven—of components (Fig. 7.6). The panel consisting of three layers of dissimilar materials enables an efficient division of functions. The low-atomic-weight beryllium or graphite is used for the plasma-facing armour, most prone to sputtering and therefore generating the low Z impurities. The massive panel permeated with water-cooling channels removing heat from the armour is made of a CuCrZr alloy, a high-plasticity and high thermal-conductivity metal. The third-layer structural material—strong

FIGURE 7.6  The first wall cross section. (1) The armour, (2) the heat sink panel, (3) the loadbearing structure.

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Fundamentals of Magnetic Thermonuclear Reactor Design

high-performance relatively cheap stainless steel—is used for the load-bearing structures, collectors, attachment and welding elements. A multilayer panel is more complex than a solid one and involves the problem of a reliable connection, but this is offset by the advantages of an efficient utilisation of different materials’ properties and functions. Generally, an engineering solution to creating a high-longevity nuclear power system, such as the ITER, is difficult to make because of the scarcity or uncertainty of initial technical information. Therefore, it has to be as variative as reasonably possible and able to accommodate potential innovations. Engineering design activities for ITER in-chamber components were initially based on these principles. At the first stage of reactor operation, the vertical target is to be coated with graphite or carbon fibre composite, CFC. These materials do not suffer from melting and are highly resistant to cracking at maximum design stresses. Their utilisation ensures the target’s efficient operation at the first stage and allows the practical issues in thermal engineering to be better defined. If these issues appear to be consistent with the use of the tungsten armour, then graphite/CFC will be replaced with tungsten before the D–T experimenting stage. We remember that the said materials are unacceptable at that stage due to the tritium accumulation risk. In addition, the need to optimise the structure and geometry of the entire divertor assembly may emerge after the first experimental phase. This may involve the adjustment of the divertor targets’ length and tilt angle and the incassette vacuum duct cross section. The ITER project provides for the divertor assembly adjustment and reconfiguration, as well as for a repeated replacement of the limiter panels, involving material and configuration changes.

7.3.2  Component Modelling: Technological and Testing Facilities A lot of joint experiments have been carried out on test benches and testing and development facilities under the ITER First Wall R&D Programme in ITER member countries. The most sizeable experimental facilities include the following: l

The SM-3 and RBT-6 reactors in Russia, used to study the effect of neutron fluence and thermal cycling damages on EU and RF FW mockups. l The JUDITH test facility in Germany, used for thermal mechanical experiments with material samples and FW component mockups, fabricated in the EU, Russia, the USA and Japan and subjected to pre-irradiation. l The IDTF (Russia) electron beam test stand with a full electron beam power of ∼0.8 MW, used for testing full-size models of targets from EU, Japan and Russia. l The VIKA and МК-200UG plasma accelerators (Russia), used to simulate plasma current disruption heat loads. An important part of the Programme is the ‘round-robin’ tests by international experts of power and temperature measurement systems on test benches in the EU, Russia, USA and Japan.

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A concise description of the testing and technological complex operated at Efremov Institute since the beginning of the 2000s is given next. The complex uses an industrialised approach to the development and testing of multilayer in-chamber components, which are part of Russia’s commitment under the ITER Joint Implementation Agreement (179 first-wall panels comprising 40% of the FW surface and all of the 60 central divertor assemblies). The complex will later be used for preinstallation testing of ITER divertor targets manufactured in Japan and the EU. The complex includes the following: l l

The TSEFEY-M and IDTF electron beam test facilities [6] (Table 7.3). Process equipment for thermal-vacuum joining of multilayer structures: vacuum brazing devices (for Be–CuCrZr Cu–CuCrZr joints) vacuum

TABLE 7.3 Characteristics of e-Beam Test Facilities Operated at Efremov Institute Parameter

TSEFEY-M

IDTF

E-beam power (kW)

20–200

80–800

Electron energy (keV)

10–40

10–60

Diameter of the heat load spot, distanced by 1 m from e-gun (mm) Min

15

40

Max

1.300

1.300

Max scanning frequency (kHz)

10

Full beam deflection angle (degree)

±40

Testing chamber inner dimensions (m) Diameter

1.5

2.2

Length

1.5

3.0

In-chamber pressure (Pa) Background pressure

<10−4

Best-estimate working pressure

∼10−2

Max working pressure

∼1.0

Water cooling system parameters: Closed loop max test pressure (MPa)

4.2

Temperature (°C)

20–130

20–150

16

30

Max pressure (MPa)

10

—

Temperature (°C)

20–650

—

Max flow rate (g/s1)

25

—

3

1

Max flow rate (m /h ) Gas cooling system parameters

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Fundamentals of Magnetic Thermonuclear Reactor Design

casting devices producing W–Cu bimetallic tiles, and diffusion bonding (HIP technology) device for CuCrZr–SS steel joints. l Special machines for electroerosion cutting of tungsten, laser beam welding and high-speed milling. l Test benches for metallography and mechanical testing. l Measuring and testing equipment (for 3D geometry, ultrasonic and X-ray control, as well as vacuum and hydraulic tests). The TSEFEY-M facility is intended for thermal engineering and material research, as well as design and process optimisation and testing of in-chamber component mockups. The TSEFEY-M experimental programme includes the following: l

l l l

l

Simulation and study of extreme quasi-stationary heat loads and their potential damaging effects on armour materials (e.g. evaporation, melting, local erosion and cracking). Search for ways to intensify the heat-exchange processes in water- and gascooled systems using one-sided surface heating. Multilayer structure evaluation in terms of strength and durability under thermal cycling loading at large temperature gradients. Optimisation of high-temperature vacuum-technology processes used at the ITER construction stage, including those involving fast temperature increase and/or decrease (e.g. degassing, thermal processing, casting, coating deposition and fast brazing). Experimental optimisation treatment and simulation testing of structures exposed to plasma-induced heat loads. The facility includes (Fig. 7.7):

l l l l l l l l

a vacuum chamber with built-in energy receiving target movable targets an electron gun with controllable beam vacuum equipment two stand-alone water-cooling circuits and a water-recirculation system closed-loop circuit for helium cooling of mockups diagnostic equipment and information gathering/processing system communications and auxiliary equipment

An electron gun (Table 7.4) mounted on the chamber upper lid along its axis provides a controllable heat load on mockups and materials. A cylindrical working chamber of around 4 m3 has ports measuring 50– 400 mm in diameter and a 700 × 800 mm2 door flange allowing to place a mockup inside the facility for testing, connect pumping equipment and make diagnostic equipment available. Mockups measuring up to 200 × 300 mm2 and weighing up to 10 kg are positioned using a movable water-cooled manipulator and are entered in through a locked chamber. Sizeable mockups, measuring 500 × 700 mm2 and weighing up to 200 kg are mounted on a rail-track target

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FIGURE 7.7  Sketch of the TSEFEY-M e-beam test facility.

TABLE 7.4 Electron Gun Key Characteristics Electron beam max current (A) Max power density on the target (kW/cm2) Acceleration voltage (kV) Acceleration voltage instability limit (%)

5 100 10–40 ≤1

bracket. Mockups measuring up to 1400 mm are delivered inside the chamber through the chamber door flange. When adjusting and testing the electron gun, a built-in energy-receiving target is employed. It absorbs up to 5 kW/cm2 thermal power and can endure 50,000 pulses. A radiation field with a given power distribution over the target surface is formed using a scanning magnetic device. A special code package enables the management and control over the exposure, positioning and sizes of geometric samples. The lowest thermal pulse length and time increment in which points are generated on the sample are 10 ms and 3.3 µs, respectively, and the maximum number of points is 2048. Turbomolecular pumps are used to evacuate the facility. Mockups, divertor qualification prototypes and functional blocks are cooled by different combinations of water supply, low-pressure water forced-flow cir-

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Fundamentals of Magnetic Thermonuclear Reactor Design

cuit (with water temperature of up to 90°C), high-pressure hot municipal distilled water forced-flow circuit (with water temperature and pressure of up to 130°C and 4.2 MPa, respectively), and a water-recirculation system. The system allows hot water-pressure to be increased pneumatically using inert gas. A highpressure helium closed loop is also available to enable on-line heat removal from gas-cooled objects under study and cooling rate control. The facility has a range of gauging and diagnostic instruments, including video cameras, infrared cameras (20°C–2000°C), pyrometers (160°C–2500°C), X-ray receivers, a multi-channel thermocouple measurement system, a massspectrometer and acoustic signal receivers (20 Hz to 25 kHz). Diagnostic equipment including X-ray sensors, thermographic cameras, pyrometers and video cameras, are mounted on the working chamber’s upper lid. There are 60 gauging channels and 16 management channels. The gauging and diagnostic system is part of a local computer network. The IDTF test bench is comparable to the TSEFEY-M facility by experimental and diagnostic capabilities, but has a much larger chamber and a much more powerful electron gun. It is used for comprehensive testing of full-scale mockups of the FW components.

7.3.3  Prevention of Destructive Events The first wall, expected to perform its functions effectively for two decades and retain its structural integrity, has to be designed with consideration for potential destructive processes. In the course of a reactor’s operation, armour tiles grow thinner due to erosion, which leads to surface temperature decrease and reduces the influx of impurities in the plasma. However, damage or deterioration may occur, raising the temperature of the tiles, which, in turn, may speed up the thermal and ion erosion processes. The most likely damages include cracking and/or detachment of the tiles, and the radiation-stimulated decrease in the armour material thermal conductivity. The development of microcracks parallel to the heat flow does not affect the heat-sinking capacity of the tiles and presents no risk. Examples include recrystallisation cracks, normal to a surface, that may result from solidification of a melted surface (Fig. 7.8AA). However, cracks that extend the full thickness of a tile (Fig. 7.8B) are a hazard: stresses arise at the tile–substrate interface due to dissimilar physical–mechanical properties of different materials. They may promote further cracking growth into the substrate material, which is unacceptable. To damp the interface stresses, it is reasonable to use a high-plasticity intermediate layer or modify the substrate using the contour profiling technique. Cracks running parallel to the tile surface decrease the tile’s thermal conductivity and increase the surface temperature (Figure 7.8C). Small cracks on individual tiles, caused by accidental defects, do not materially affect the aggregate flow of impurities. However, an en masse cracking, caused, for example, by

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FIGURE 7.8  Types of thermal mechanical defects occurring in armour tiles.

an incorrect machining of tile sides, may lead to a considerable contamination and cooling of the plasma. An extreme case of side cracking is a complete or partial chipping of a tile (Fig. 7.8D and E). If this defect is accidental and the affected tile does not face the plasma directly, the flaked-off piece is heated and evaporated slowly, without affecting the plasma behaviour. But if the tile is from the upper reactor part, a chipped off piece of tile may get into the plasma column and cause a current disruption. If chipping is intense or affects tiles that face the plasma directly, the reactor may go out of service. The detachment of an entire tile is dangerous not only because the tile can get inside the plasma (which in some cases is tolerable). The greatest risk is the substrate’s lengthy exposure to the plasma particle fluxes and potential leakage of copper into the plasma column. Let us make some quantitative estimates. We start with assuming that bremsstrahlung radiation accounts for the largest part of plasma energy losses. Bremsstrahlung radiation power is proportional to Z2. Next, we assume that beryllium, graphite and copper have commensurate sputtering coefficients. With these assumptions in hand, it is clear that one can ‘open’ just 4% of the heat sink panels’ surface area (∼25 m2) to double the theoretical radiative losses. If we use tighter criteria and assume that the losses are mostly caused by recombination radiation with power proportional to Z4, then the ‘allowable’ loss

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of armour will be ∼1 m2. A tile measures 2.5 × 10−4 m2 at most. Then, the above observation that the loss of one tile is allowable is correct, but to the extent that the loss of one tile and the resultant heat flow redistribution do not induce avalanche-like damage of neighbouring tiles. As a result of irregularities occurring due to tile partial detachment or deformation at the tile–plasma interface, prominent parts become overheated, which entails a series of damages, as shown in Fig. 7.8G and H). Such irregularities are allowable if there is no direct contact with the plasma. For plasma-facing FW components, the largest permissible vertical displacement of one tile against another must be within 0.3 mm, while the difference in height between neighbouring ICCs may be up to 3 mm. Another thing worthy of consideration is the impact produced by the armour on the heat-sink panels’ operating conditions. Transient events, such as current disruptions, may make the panels exposed to extreme heat fluxes with resultant overheating and even melting of their surfaces. This, in turn, may initiate surface cracking. However, ITER design requirements forbid any cracks on any structural components containing a forced-flow cooling system. The width-toheight ratio must be as small as technologically permissible to decrease the heat loads on the nonarmoured part of panels (Fig. 7.8F). The design and technological solutions for the heat sink panels, load-bearing structures and collectors must also meet requirements, such as the ‘absolute’ leak-tightness (total number of uncontrollable gas leaks through the many-kilometre-long welding joints and cooling channels must be within 1.10−8 Pа m3/s. In-chamber components must not experience deformations that may give rise to irregularities and local curvatures of plasma-facing surfaces. Fastener assemblies must be designed with online remote replacement capability. Despite the radiation-induced degradation of the piping material, the opportunity to hermetically weld on replaceable ICCs from time to time must be provided. The difficulties listed emphasise the need to comprehensively evaluate and analyse the design solutions from the thermal mechanics perspective.

7.4  NEXT-GENERATION REACTOR FIRST WALL 7.4.1 Challenges The engineering ideas implemented in the ITER functional systems have raised reactor-making to the heights of technical mastery. However, the ITER is just an experimental machine, and its characteristics are far from the projected parameters of a demonstrational reactor [7] to say nothing about a commercial power reactor! We further make some simple comparisons (Table 7.5). The plasma column sizes and divertor target areas being similar, the DEMO reactor plasma will be hotter. In addition, the DEMO’s first wall will be exposed to a heat flow of much greater strength and density, while the cycle active phase will be many times

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TABLE 7.5 Tokamak Divertor Operating Conditions (Project Evaluations) Reactor Parameter

ITER

DEMO

Heat flow (MW)

∼100

>300

On sloped target

20

∼60

In divertor plasma, max

∼500

∼2000

Pulse length(s)

400

>104

Heat flow density (MW/m2)

Number of pulses

310

4

Plasma total burn time during operation cycle(s)

∼10

Plasma relative burn time during operation cycle (%)

4

∼104

7

>108 up to 40 3

Number of current disruptions

3 × 10 (10%)

≤102 (1%)

Expected number of target replacements

≥5

up to 2

Radiation-induced material degradation over reactor operation time (dpa)

∼1

≥30

Dust generation (kg)

100

>1000

Materials relative activation over reactor operation time

1

>30

Helium evacuation flow (Pa m3/s)

0.7

>2.0

longer. Given these data, the theoretical erosion lifetime of DEMO ‘classical’ divertor targets will be at least 10× lower, by optimistic estimates. As the targetreplacement outages are bound to affect DEMO’s economic performance, the unambiguous conclusion is that the conventional FW in the form of a static water-cooled configuration with periodically replaceable elements will not do the trick in the next-generation reactors. A new conceptual approach is needed to achieve a few orders of magnitude increase in the divertor target lifetime and improve the heat sinking capability. Also, it is necessary to increase the coolant’s temperature, helium evacuation efficiency, the components’ radiation resistance, and radically decrease dust generation.

7.4.2  Possible Engineering and Physical Solutions Let us first consider the erosion lifetime issue. There may be three possible ways to address it: (1) suppress erosion (rate) by physical and technological methods, (2) build up the sacrificial material initial mass and (3) apply additional sacrificial layer during reactor operation (Table 7.6). One well-established way to increase the erosion lifetime is to use an armour material with the lowest possible sputtering coefficient, such as tungsten, and

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TABLE 7.6 Methods for Increasing the Erosion Lifetime of Divertor Targets Concept

Method

Technical solutions

Erosion rate minimisation

Material selection Optimisation of edge plasma parameters

Tungsten armour Edge plasma cooling down and reduction of oxygen content

Increase of initial mass of the eroding material

Erosion area expansion Material re-usage Building up the sacrificial layer

Target movement Separatrix displacement Rotating target Re-melted target Use of first wall material Tile design optimisation Multi-channel target

Addition of fresh material

Restoring deposition Facilitate replacement of armour/target Continuously renewed armour

Plasma feeding Chemical vapour deposition Replacement of tiles/target Liquid metal films Liquid metal jets Liquid or solid droplet curtain Evaporating target with capillary substrate Plasma spray Thermal evaporation/deposition

establish physical conditions at the plasma periphery (low-edge plasma’s temperature and reduced oxygen impurity concentration) such that tungsten could present itself to its best advantage. The sacrificial material’s initial mass can be increased by spreading the plasma particle flux over a larger area, providing a quasiclosed atom circulation cycle (sputtering–redeposition), and building up the sacrificial layer. Eroded surface extension is achieved by displacing the target and the peak of a separatrix parallel to the plasma particle flow against each other (Fig. 7.3). To this end, either rotating targets (Fig. 7.9) or moving the separatrix by poloidal magnetic field variation are employed. The rate of such relative displacement must be optimised (see comment to Table 7.7). Sputtered atoms deposit on the FW surface parts, where erosion is slow or absent. They can be driven back to intense-erosion regions by aligning the axes of rotating targets along the toroidal magnetic field direction. In this case, the concentration of the ‘returnee’ atoms over the target surfaces will be more uniform. The initial topography of the sputtered surface can also be restored by thermo-gravitational methods that get a layer of deposited atoms melted and then redeposited under gravity (Fig. 7.10). The possibilities of increasing the sacrificial layer by optimising the combinations of properties related to heat transfer (thermal conductivity—the largest permissible surface temperature) have largely been exhausted. One opportunity that remains unexploited is the multichannel target that employs consecutive switching of the coolant path (Fig. 7.11).

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FIGURE 7.9  Possible configurations including rotating cylinders.

TABLE 7.7 Methods for Increasing Heat Loads, Which Divertor Targets Can Steadily Withstand Largest permissible heat flow density (MW/m2)

Concept

Method

Expansion of target area withstanding the heat flow

Separatrix repositioning Target displacement

Heat flows partial redistribution

Use of radiating target Particle reflection Use of evaporating target

2.5

Cooling mode optimisation

Enhancement of coolant’s heat absorbing capacity Heat exchange intensification

Water (40.0) Helium (10.0) Liquid metal pumped through the channel (5.0) Freely flowing liquid metal (30.0)

40.0

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Fundamentals of Magnetic Thermonuclear Reactor Design

FIGURE 7.10  Bath with a melted filler.

FIGURE 7.11  The multichannel target.

FIGURE 7.12  Swinging target with a liquid-metal interlayer.

An alternative way to improve the erosion lifetime is to replace eroded material with a ‘new’ one. This can be done using one of the following engineering techniques [8–10]: l

An in situ deposition of new layers on areas subject to particularly strong erosion using built-in evaporators or sacrificial material introduced in edge plasma.

l

An express replacement of tiles or target parts. This is done using the simplest ‘armour–heat sink panel’ separable connection in the form of a liquidmetal interlayer (Fig. 7.12) or a radiative connector based on a heat pipe (Fig. 7.13). l Continuous surface restoration with liquid metal (Fig. 7.14). l Use of a solid ball shower curtain.

FIGURE 7.13  Heat pipe-based swinging target.

FIGURE 7.14  Different liquid-metal targets.

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Fundamentals of Magnetic Thermonuclear Reactor Design

Similar approaches can be used to intensify heat removal from divertor targets (Table 7.7). Reciprocation of a strongly peaked heat flow over a target surface can be performed in two ways: (1) using movable targets and (2) repositioning the separatrix through variations of the poloidal field. To achieve an appreciable positive result, the amplitude of the separatrix and target displacement against each other must be larger than the heat flow’s azimuthal ‘length’ (0.1–0.3 m), with the least required displacement frequency close to 0.3 Hz. The heat load can be dispersed by a hot target, but even at the highest tolerable temperature of ∼2500°C not more than 2.5 MW/m2 can be radiated, a number negligible compared with the ∼40 MW/m2 flow incident on the target. For a flow coming at a small angle of incidence, the power reflection coefficient may reach ∼50%. In divertor targets, the glancing angle of the incident plasma flow is indeed very small (in the order of several degrees). However, this reflection coefficient is impossible to achieve in practice. The reason is the induced electric potential that develops near the surface and makes reflected ions return to the target. Therefore, the best method to effectively withstand the severe heat loads in a stationary environment would be to evaporate the sacrificial layer of the target. For example, the evaporation of a liquid lithium coating at 1200°C– 1500°C enables the absorption of a heat flow of up to 50 MW/m2. However, this method has a fundamental weakness: plasma contamination by evaporating lithium. Another method for intensifying heat transfer is the cooling mode optimisation. This can be achieved by improving the coolant’s thermal accumulation capacity at the wall–coolant interface. Factors to be controlled include the coolant itself, the coolant mass flow rate, the coolant in-channel warming up, the channels’ inner microrelief, the flow turbulence degree, and the coolant’s phase transitions. Coolants employed in present-day power engineering have different heat accumulation capacities. For example, water flowing at the highest permissible rate of 10–15 m/s can withstand stationary loads of up 40 MW/m2. This is close to loads expected in the MFR, but provides no margin. In addition, thermal engineering systems using a water coolant have a relatively low heat-to-electric energy conversion coefficient. Many commercial cooling systems utilise gaseous helium which, with all its operational merits, has a very low heat accumulation capacity. At a practically maximum pressure of 20 MPa and a flow rate of up to 100 m/s, the highest heat load on a helium-based cooling system is within 10–15 MW/m2. Liquid metals as a specific class of coolants have important thermal engineering advantages, but when employed in an MFR, they feature limitations associated with magnetic fields that constrain the permissible pumping speed. The liquid metals’ highest permissible speed of circulation in a closed-circuit cooling system is close to 1 m/s, while the largest permissible heat load is

TABLE 7.8 Parameters and Characteristics of Energy-Intensive Components of Nuclear Facilities Parameter/Characteristic

MFR first wall

Nuclear fuel elements

Particle accelerator targets

∼103

∼103

∼10−2

Heat flow density limit (MW/m )

∼30

∼1,5

∼10

Ion erosion

++

−

+

Neutron-induced damages (dpa)

10

∼10

0.1–10

∼10

Up to 104

Structural system

Typical surface area (m2) 2

Number of thermal cycles

Up to 10

Operation life (h)

Up to 104

Up to 105

Up to 103

Coolant

Water, helium and liquid metal

Water

Water

Armour

Be, C, W

—

Be, C, W Cu

Panel

Cu and stainless steel

Zr

Operating environment

High vacuum and hydrogen

Non-vacuum

Material

High vacuum

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4

243

Vacuum tube electrodes

High power laser mirrors

Rocket nozzles

Aircraft/spacecraft fairings

Steam generators

∼1

∼1

∼10

∼100

∼1000

Heat flow density limit (MW/m )

∼10

up to 10

up to 30

up to 5

0.3

Number of thermal cycles

Up to 106

Up to 105

1

Up to 102

Up to 104

Operation life (h)

Up to 104

∼104

∼0.1

Up to 10

∼105

Coolant

Water

Water

—

—

Vapour-water mixture

Armour

Cu, Ta, W

SiC, Cu

C–C composite, W

C–C composite, SiC, SiO2

Cu, Cu–Ni

Panel

Cu

Ве, SiC, Cu

—

—

Cu–Ag, steel

Operating environment

High vacuum

Atmospheric air

Atmospheric air, oxidising agents

Atmospheric, ionospheric air

Atmospheric air

Parameter/Characteristic

Area (m2) 2

Material

Fundamentals of Magnetic Thermonuclear Reactor Design

Component

244

TABLE 7.9 Parameters and Characteristics of Components of Thermal and Power Engineering Systems and Facilities

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FIGURE 7.15  Mockups of Be-targets of an electro-nuclear neutron generator.

within 10 MW/m2. When operated in a free-flow manner, liquid metals (films, jets and drops) may reach up to ∼10 m/s and withstand heat loads of up to ∼30 MW/m2. In the midterm, some of the present-day ideas, concepts and approaches may prove unviable and give way to new, currently unbeknown to us, solutions.

7.5  ALTERNATIVE USES OF FIRST-WALL TECHNOLOGIES As one can see from Tables 7.8 and 7.9, the MFR’s first wall has a number of features that place it into more or less one class with energy-intensive systems belonging to other branches of engineering. For this reason, processes developed for the first wall may be suitable for other applications. One example is the target of an electro-nuclear neutron generator. In this device, a water-cooled beryllium target is bombarded with accelerated protons, giving rise to fast neutrons. The operating conditions of the target are similar to those of the first wall and include a cyclic heat load of up to 5 MW/m2, high vacuum and radiation-induced damages. The specimens shown in Fig. 7.15 have successfully passed endurance tests. Going forward, the first-wall technologies may be utilised in other machines and systems operating under severe heat conditions.

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