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Neutral beam deployment on DEMO and its influence on design
Fusion Engineering and Design 86 (2011) 793–796
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Neutral beam deployment on DEMO and its influence on design Elizabeth Surrey ∗ , Damian King, Jonathan Lister, Michael Porton, William Timmis, David Ward EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, United Kingdom
a r t i c l e
i n f o
Article history: Available online 26 May 2011 Keywords: Neutral beam injection Heating and current drive DEMO Negative ion
a b s t r a c t The demands on the neutral beam heating and current drive system of a DEMO device exceed those of existing fusion experiments by several orders of magnitude. By predicting possible power waveforms it is possible to analyse the technological advances necessary to achieve a system relevant to deployment on a power plant. Achieving the necessary efficiency will require simultaneous improvements in beam current density, neutralization efficiency and beam transmission. Considering the deployment on the tokamak vessel shows no major disruption to the tritium breeder blanket and no requirement to reach a high packing density of injectors. The thermal management of components subjected to low heat flux for many hours is considered and it is shown that radiation cooling can be exploited to control the temperature of such items. © 2011 EURATOM Culham Centre for Fusion Energy. Published by Elsevier B.V. All rights reserved.
1. Introduction The development of nuclear fusion beyond the advanced experimental phase represented by ITER requires the construction of a quasi-power plant demonstrator commonly referred to as DEMO. Several versions of the DEMO concept exist, varying in size and modus operandi and each variant places different demands on the Heating & Current Drive (H&CD) systems. The power and pulse length requirements in particular influence the performance and design of these systems and this is most apparent for neutral beam injection (NBI). The various options for NBI are examined and their influence on aspects of deployment on the tokamak and design of the beamline are discussed. Finally, areas requiring significant R&D are identified. 2. Determination of DEMO NBI requirements A number of power plant studies [1–4] have produced a range of operating conditions for a DEMO device, from steady state to pulsed plasma operation. The purpose of NBI also varies between designs. For example, the steady state of [2] requires ∼240 MW of current drive (CD), most of which will be provided by NBI due to the relatively low efficiency of other systems [5], whereas the pulsed option of [3] has no CD requirements. Steady state performance will also require NBI to perform some degree of real time plasma control; possibly burn control or current profile control. This implies that a degree of modulation of the NB power will be necessary. For burn
∗ Corresponding author at: CCFE, Culham Science Centre, Abingdon OX14 3DB, United Kingdom. Tel.: +44 1235 464473. E-mail address: [email protected] (E. Surrey).
control, relatively low power (∼25 MW) at a frequency of ∼0.05 Hz is indicated in [6] and for current profile ∼20% of total NBI power for several seconds is indicated in [7]. The power waveforms for DEMO are as yet undefined but a nominal 10% modulation will be assumed, implying that the minimum number of injector units for the steady state device is ten. Another feature of the steady state DEMO is the NBI pulse length of 10–240 h. All DEMO devices will require ∼100 MW NBI power for 10 min to reach the burn phase followed by a further 20 min of additional heating at 50 MW for the pulsed DEMO option. These pulse lengths are considerably longer than those on existing tokamaks and it is shown in Section 3.4 that there are consequences for the design of the beamlines. After these periods of operation, the NBI systems may be turned off for periods of hours before restarting, so the continuous cycling will result in creep-fatigue of the components. The beam energy is determined by the plasma density and electron temperature, which dictate the beam energy loss rate, and the plasma minor radius which dictates the depth of penetration required. Most DEMO designs have plasma radii ∼2.4 m and for an average density of 1020 m−3 and temperature 16 keV penetration to this depth requires beam energies above 500 keV/amu. We will assume a neutral beam energy of 1.5 MeV, derived from a negative ion deuterium precursor beam.
3. Analysis of NBI system The key factors in determining the NBI system deployment on a DEMO device are beam current density, neutralization efficiency and beam transmission. By analysis all other parameters can be shown to determine these quantities. For example, the beam divergence and uniformity are simply quantities that determine the
0920-3796/$ – see front matter © 2011 EURATOM Culham Centre for Fusion Energy. Published by Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.01.021
794
E. Surrey et al. / Fusion Engineering and Design 86 (2011) 793–796
Table 1 NB technology options for configuration analysis. Beamline
A
B
C
D
E
Current Density (A m−2 ) Injector power (MW) Neutralization efficiency Transmission Core divergence (mrad) Halo divergence (mrad) Electrical efficiency (%)
260 9 0.58 0.75 7 15 27
260 9 0.58 0.95 3 6 34
390 13.5 0.58 0.75 7 15 28
260 9 0.95 0.85 7 15 52
390 13.5 0.95 0.95 3 6 65
transmission. The status of these fundamental quantities is represented by the ITER design for the HNB given in the ITER Design Document [8]. The ITER D− source current density of 260 A m−2 has been demonstrated in small (JET PINI sized) sources to date and for relatively short pulse lengths (100 s) [9]. The actual beam divergence is unknown so the ITER DDD approaches the problem by adopting three possible core divergences 3 mrad, 5 mrad and 7 mrad accompanied by a beam halo, containing 15% of the beam particles, characterized by a divergence approximately twice that of the core. The core divergence is a critical quantity for determining the beam transmission. Using the Beam Transmission & Reionization code [8] the transmission to the tokamak is shown to vary from 70% to 95% over this range of divergence. The neutralization efficiency of the beam for a gas neutralizer is 58% but can be increased to 95% by adopting the photoneutralizer [10]. Interest in this option has been driven by the need to improve the electrical efficiency of the NBI system to around 60%. Five beamlines have been considered each representing a different combination of these key factors as defined in Table 1. The implication of deploying these beamlines on a DEMO device can then be assessed to prioritize future R&D effort. There is no reason to assume large single, injectors (as on ITER) will occupy less space in the blanket than multiple smaller units linked through a common aperture and the latter arrangement offers reduced impact in the event of a unit failing. We have therefore adopted a basic injector unit of 6A D− giving 9 MW negative ion beam power (beamline A). This corresponds to extraction from a JET-PINI sized source [11] with the ITER current density. Another alternative, the lithium vapour jet, with neutralization efficiency of 63% corresponding to electrical efficiency of 35–40% has not been considered. 3.1. Injector configuration The steady state modulated waveform is the most demanding and Fig. 1 shows the number of beam ports required to deliver 240 MW from each of the beamline options in Table 1 as a function of the number of units per port. There is little to be gained by
striving for high packing density as the relative reduction in port area (and hence interruption to the tritium breeder blanket) is lessened. The power density per port varies from 40 to 100 MW m−2 (for comparison the JET port transmits 46.2 MW m−2 ). Choosing a conservative option of beamline A, the required CD power could be delivered through 16 ports with 4 injectors per port, a total area of 4 m2 , less than 1% of the surface area of the tokamak. The greatest single benefit is to be gained by improving the neutralization efficiency to 95% through the photoneutralizer (D). However given the challenging nature of this development other routes to improvement are considered. Increasing the current density (C) would enable either the number of ports or the number of injectors per unit to be reduced. However, if the packing density is high (say 6 units or more) the differential between an increase in current density (C) and an increase in transmission (B) is reduced. The presence of beamline components, such as calorimeters can have a deleterious influence on the beam transmission. For example, in beamline (A) most of the transmission losses occur in the neutralizer and residual ion dump channels, so there is obviously scope for improvements in design of the beamline. This may be easier to achieve than an increase in current density, thus favouring a high packing density. Conversely, selecting fewer, larger sources (say 4 per beamline) forces development along the path of increased current density as the benefits of increased transmission are depleted. Although increasing the source current density reduces the number of units required, it is clear from Section 3.4 that any such increase should be accompanied by an improvement in the transmission or, more specifically, a reduction in direct interception on beamline components. This is desirable even without an increase in current density if the multi-hour, long pulse mode is to be realised. There is, however, advantage to reducing the number of injectors per unit as it allows greater freedom in determining HV stand-off distances between injectors. To compensate the injector size could increase. Large area sources also yield a higher fraction of extraction area per unit footprint area, when the HV standoff distance is taken into account and in this respect sources with an almost square profile are more efficient than rectangular. One disadvantage of large sources is the degree of steering of the beams into the port aperture that can be achieved. On JET, for instance, the individual injector units are physically oriented at angles of 158 mrad and 51 mrad to the horizontal plane and 61 mrad to the vertical plane in addition to grid inclination and offset aperture steering. On ITER the beam steering is limited to 16 mrad through grid and offset aperture and this contributes to the relatively large port size. Offset aperture steering depends on the magnitude of the electric field in the accelerator and is limited by the onset of beam aberration and direct interception on the grid. It is unlikely that steering greater than that adopted for ITER would be acceptable. It is clear that the only significant improvement is through a combination of all three strategies (E), i.e. increasing current density, improving transmission and increasing the neutralization efficiency. This represents a significant R&D effort over the next two decades. 3.2. Electrical efficiency
Fig. 1. Number of ports required to inject a total of 240 MW for different beamline technologies in Table 1.
Again, the steady state waveform is the critical case; with the heating waveforms the beams are only required for a small fraction of the plasma cycle and so their impact on overall efficiency is reduced. The electrical efficiency for each beamline combination is given in Table 1; the efficiency is estimated by including efficiencies for power supplies and support services such as cryoplant. As before, the single most influential improvement to be made is in the neutralization efficiency. Even if the photoneutralizer were to
E. Surrey et al. / Fusion Engineering and Design 86 (2011) 793–796
795
Fig. 2. Relationship between power per beamline and pulse length for existing fusion experiments and the future JET EP2, ITER and DEMO devices. Note uncertainty on DEMO power indicated by error bars.
prove possible, achieving the desired electrical efficiency of 60% still requires optimistic and simultaneous scaling of the other factors. This is a key point for the deployment on NBI on DEMO: increasing the NBI efficiency from 27% to 60% reduces the installed NB power from ∼890 MW to ∼400 MW (to deliver 240 MW). At 27% NBI efficiency the power output from a 2.5 GW (thermal) reactor with generating efficiency of 40% is almost entirely consumed by the NBI system, necessitating a much larger reactor. 3.3. Caesium consumption The ITER source uses caesium to enhance the surface production of negative ions to produce the required density. The PINI-sized prototype source consumes 14 g of caesium per second of plasma time [9]. It is not clear at present if the caesium consumption is dominated by diffusion out of the source or by gettering at the wall but as the extraction area and source wall area are closely linked in volume sources (at least to date) either mechanism can be used to scale for a DEMO requirement of ∼24 g caesium per 10 h pulse or ∼17 kg per year. Apart from the obvious safety issue of storing such quantities of caesium, some of the metal will diffuse into the accelerator and cause breakdown of insulators. It is clear that an alternative to caesium would be highly desirable and this may require adopting alternative source technology such as the pulsed afterglow plasma [12] and this may require a change in the philosophy of neutral beam injection. 3.4. Long pulse operation 3.4.1. Thermal management The anticipated pulse length of the steady state waveform, upwards of 10 h, is another significant departure from present experience with NB systems. The extent to which DEMO stretches the operational envelope is illustrated in Fig. 2 where the NB power per beamline (or port) is plotted as a function of the pulse length for a number of fusion experiments. The inverse nature of the power transmitted through the beamline and the pulse length in most current experiments is clear and arises from the engineering design of the beamline and its components. The location of the ITER operating space and possible DEMO parameters lie significantly outside current experience and will provide particular challenges in engineering design and material science. Of concern are the many un-cooled components, particularly at the periphery of the beamline, where the incident heat flux is generally regarded as trivial and so are not subject to full design scrutiny. To investigate the effect of low heat flux over a prolonged period a finite element model of a generic component, (a CuCrZr
Fig. 3. Finite element model of generic peripheral beamline component used in long pulse study.
plate bolted to a steel vessel with an average contact pressure of 2.7 M N m−2 , giving a thermal conductivity of 700 W m−2 K−1 for a surface finish of 1–1.5 m [13]), typical of those used for protection on current fusion devices, was constructed. In the model there is no active cooling in the region, and the only available mechanism of external cooling is through natural air convection at the top face of the vessel (i.e. radiation is ignored). The equilibrium temperature distribution (reached in ∼20 h), calculated for a relatively low value steady state incident heat flux of 5 kW m−2 , such as might be caused by interception of stray beam ions, is illustrated in Fig. 3 where a peak temperature of 1200 K is obtained for the most severe conditions considered. Increasing the conductivity at the interface to infinity only reduces the maximum temperature by 6% but a tenfold increase of the convection coefficient at the outer vessel wall reduces the peak temperature by one third. Regarding radiation, we consider the generic component to be enclosed by a surface of finite emissivity and area. Thus the enclosure can also radiate heat back to the component, a process that is characterized by the geometric viewing factor, emissivities, areas and temperatures of the component and the enclosure. The effect of varying the emissivity of either the component or the enclosure and of the viewing factor is shown in Fig. 4 for the conditions of Fig. 3 and equal radiative areas of component and enclosure. There is no preferential effect between the emissivities; higher values lead to lower component temperature, although the effect diminishes above 0.5. Similarly the viewing factor has the largest impact between 0 and 0.5. (The temperature of the enclosure does not have a significant impact below 450 K.) Thus by manipulating the emissivity of the component and enclosure surface and by geometric consideration of the viewing factor, thermal management of peripheral components should be feasible without resorting to active cooling. 3.4.2. Mechanical failure In addition to the thermal aspects of long pulse operation discussed in Section 3.4.1, there are two mechanical failure mechanisms associated with long pulse operation: creep-fatigue and irradiation creep. Creep-fatigue is a combination of normal creep mechanism encountered in continuous operation and fatigue encountered in low duty cycle operation. The NBI system will be subject to cyclic stress superposed on a constant load as a result of either modula-
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E. Surrey et al. / Fusion Engineering and Design 86 (2011) 793–796
JG10.22-4c
component temperature (K)
1200 1100 1000 900 800 700 600
0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1
emissivity or view factor Fig. 4. Effect on component temperature of variation in emissivity, ε and view factor: component varies enclosure ε = 0.5, enclosure varies, component ε = 0.5, enclosure varies, component ε = 0.25 and solid line view factor varies (both ε = 0.5).
tion in the steady state DEMO or the cyclic waveform for the heating phase of the pulsed DEMO. The combination lowers the lifetime of components significantly; a reduction of almost 50% was recorded by Wu et al. [14] and the effect is temperature dependent [15]. Irradiation creep arises from displacement damage due to neutrons and the production of transmutant helium. The effect is sensitive to temperature and neutron energy and has been studied extensively for the fast fission neutron spectrum but not for the fusion spectrum. A definitive description of irradiation creep combined with creep-fatigue is given in [16], where an example for Austenitic 316 steel shows a reduction in lifetime of an order of magnitude compared to thermal creep only. Current design codes do not address irradiation, although the methodology of [16] could be used if more fusion relevant data were available. 4. Conclusions The possible operational demands on NBI of a DEMO device have been assessed in terms of NB power and pulse length. Both are outside present operational experience and represent significant challenges to the technology. Three key parameters are identified for future development: beam current density, neutralization effi-
ciency and beam transmission and it is shown that improvements in all three must be simultaneously achieved to obtain an electrical efficiency relevant to power generating plant. The requirement for NBI power modulation precludes using a small number of large injectors although this approach is more efficient in terms of spatial use and in this respect square profile sources make the most efficient use of the stacking footprint area. Considering the deployment of the NBI system on the tokamak, there is little point in striving for high packing density per port as the interruption to the blanket is less than 1% of total area. Finally, consideration has been given to the effect of low heat flux acting over prolonged periods on uncooled components, such as may be found in the periphery of the beamline. It is shown that by manipulating the emissivity of the component and enclosure surfaces and by implementing the correct geometry with respect to their enclosure, the temperature of these items can be controlled through radiation cooling alone. Acknowledgements This work was funded by the United Kingdom Engineering and Physical Sciences Research Council under grant EP/G003955 and the European Communities under the contract of Association between EURATOM and CCFE. The views and opinions expressed herein do not necessarily reflect those of the European Commission. References [1] D. Maisonnier, et al., Nucl. Fus. 47 (2007) 1524. [2] D.J. Ward, W.E. Han, Results of System Studies for DEMO, TW6-TRP-002 CCFE, 2007. [3] G.M. Voss, et al., Fus. Eng. Des. 63–64 (2002) 65. [4] F. Najmabadi, et al., Fus. Eng. Des. 80 (2006) 3. [5] D.J. Campbell, et al., IAEA-CN-FT/1-1, in: 21st Fus. Energy Conf., Chengdu, China, 2006. [6] H.P.L. de Esch, et al., Fus. Eng. Des. 26 (1995) 589. [7] T. Suzuki, et al., Nucl. Fus. 48 (2008) 045002. [8] ITER DDD 5.3, N53 DDD 29 01-07-03 R0.1 IAEA, Vienna (2003). [9] W. Kraus, et al., 1st Symp. Neg. Ions, Beam & Sources Aix-en-Provence 2008, in: AIP Conf. Proc., vol. 1097, 2009, p. 275. [10] M. Kovari, B. Crowley, Fus. Eng. Des. 85 (2010) 745. [11] G. Duesing, et al., Fus. Technol. 11 (1987) 163. [12] M.B. Hopkins, et al., J. Appl. Phys. 70 (2009) 1991. [13] W. Rohsenhow, Handbook of Heat Transfer Fundamentals, McGray Hill, 1973. [14] X. Wu, et al., J. Nucl. Mater. 367–370 (2007) 984. [15] S.D. Preston, et al., Fus. Eng. Des. 48 (2003) 527. [16] P.J. Karditsas, Fus. Eng. Des. 48 (2000) 527.
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Neutral beam deployment on DEMO and its influence on design Elizabeth Surrey ∗ , Damian King, Jonathan Lister, Michael Porton, William Timmis, David Ward EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB, United Kingdom
a r t i c l e
i n f o
Article history: Available online 26 May 2011 Keywords: Neutral beam injection Heating and current drive DEMO Negative ion
a b s t r a c t The demands on the neutral beam heating and current drive system of a DEMO device exceed those of existing fusion experiments by several orders of magnitude. By predicting possible power waveforms it is possible to analyse the technological advances necessary to achieve a system relevant to deployment on a power plant. Achieving the necessary efficiency will require simultaneous improvements in beam current density, neutralization efficiency and beam transmission. Considering the deployment on the tokamak vessel shows no major disruption to the tritium breeder blanket and no requirement to reach a high packing density of injectors. The thermal management of components subjected to low heat flux for many hours is considered and it is shown that radiation cooling can be exploited to control the temperature of such items. © 2011 EURATOM Culham Centre for Fusion Energy. Published by Elsevier B.V. All rights reserved.
1. Introduction The development of nuclear fusion beyond the advanced experimental phase represented by ITER requires the construction of a quasi-power plant demonstrator commonly referred to as DEMO. Several versions of the DEMO concept exist, varying in size and modus operandi and each variant places different demands on the Heating & Current Drive (H&CD) systems. The power and pulse length requirements in particular influence the performance and design of these systems and this is most apparent for neutral beam injection (NBI). The various options for NBI are examined and their influence on aspects of deployment on the tokamak and design of the beamline are discussed. Finally, areas requiring significant R&D are identified. 2. Determination of DEMO NBI requirements A number of power plant studies [1–4] have produced a range of operating conditions for a DEMO device, from steady state to pulsed plasma operation. The purpose of NBI also varies between designs. For example, the steady state of [2] requires ∼240 MW of current drive (CD), most of which will be provided by NBI due to the relatively low efficiency of other systems [5], whereas the pulsed option of [3] has no CD requirements. Steady state performance will also require NBI to perform some degree of real time plasma control; possibly burn control or current profile control. This implies that a degree of modulation of the NB power will be necessary. For burn
∗ Corresponding author at: CCFE, Culham Science Centre, Abingdon OX14 3DB, United Kingdom. Tel.: +44 1235 464473. E-mail address: [email protected] (E. Surrey).
control, relatively low power (∼25 MW) at a frequency of ∼0.05 Hz is indicated in [6] and for current profile ∼20% of total NBI power for several seconds is indicated in [7]. The power waveforms for DEMO are as yet undefined but a nominal 10% modulation will be assumed, implying that the minimum number of injector units for the steady state device is ten. Another feature of the steady state DEMO is the NBI pulse length of 10–240 h. All DEMO devices will require ∼100 MW NBI power for 10 min to reach the burn phase followed by a further 20 min of additional heating at 50 MW for the pulsed DEMO option. These pulse lengths are considerably longer than those on existing tokamaks and it is shown in Section 3.4 that there are consequences for the design of the beamlines. After these periods of operation, the NBI systems may be turned off for periods of hours before restarting, so the continuous cycling will result in creep-fatigue of the components. The beam energy is determined by the plasma density and electron temperature, which dictate the beam energy loss rate, and the plasma minor radius which dictates the depth of penetration required. Most DEMO designs have plasma radii ∼2.4 m and for an average density of 1020 m−3 and temperature 16 keV penetration to this depth requires beam energies above 500 keV/amu. We will assume a neutral beam energy of 1.5 MeV, derived from a negative ion deuterium precursor beam.
3. Analysis of NBI system The key factors in determining the NBI system deployment on a DEMO device are beam current density, neutralization efficiency and beam transmission. By analysis all other parameters can be shown to determine these quantities. For example, the beam divergence and uniformity are simply quantities that determine the
0920-3796/$ – see front matter © 2011 EURATOM Culham Centre for Fusion Energy. Published by Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2011.01.021
794
E. Surrey et al. / Fusion Engineering and Design 86 (2011) 793–796
Table 1 NB technology options for configuration analysis. Beamline
A
B
C
D
E
Current Density (A m−2 ) Injector power (MW) Neutralization efficiency Transmission Core divergence (mrad) Halo divergence (mrad) Electrical efficiency (%)
260 9 0.58 0.75 7 15 27
260 9 0.58 0.95 3 6 34
390 13.5 0.58 0.75 7 15 28
260 9 0.95 0.85 7 15 52
390 13.5 0.95 0.95 3 6 65
transmission. The status of these fundamental quantities is represented by the ITER design for the HNB given in the ITER Design Document [8]. The ITER D− source current density of 260 A m−2 has been demonstrated in small (JET PINI sized) sources to date and for relatively short pulse lengths (100 s) [9]. The actual beam divergence is unknown so the ITER DDD approaches the problem by adopting three possible core divergences 3 mrad, 5 mrad and 7 mrad accompanied by a beam halo, containing 15% of the beam particles, characterized by a divergence approximately twice that of the core. The core divergence is a critical quantity for determining the beam transmission. Using the Beam Transmission & Reionization code [8] the transmission to the tokamak is shown to vary from 70% to 95% over this range of divergence. The neutralization efficiency of the beam for a gas neutralizer is 58% but can be increased to 95% by adopting the photoneutralizer [10]. Interest in this option has been driven by the need to improve the electrical efficiency of the NBI system to around 60%. Five beamlines have been considered each representing a different combination of these key factors as defined in Table 1. The implication of deploying these beamlines on a DEMO device can then be assessed to prioritize future R&D effort. There is no reason to assume large single, injectors (as on ITER) will occupy less space in the blanket than multiple smaller units linked through a common aperture and the latter arrangement offers reduced impact in the event of a unit failing. We have therefore adopted a basic injector unit of 6A D− giving 9 MW negative ion beam power (beamline A). This corresponds to extraction from a JET-PINI sized source [11] with the ITER current density. Another alternative, the lithium vapour jet, with neutralization efficiency of 63% corresponding to electrical efficiency of 35–40% has not been considered. 3.1. Injector configuration The steady state modulated waveform is the most demanding and Fig. 1 shows the number of beam ports required to deliver 240 MW from each of the beamline options in Table 1 as a function of the number of units per port. There is little to be gained by
striving for high packing density as the relative reduction in port area (and hence interruption to the tritium breeder blanket) is lessened. The power density per port varies from 40 to 100 MW m−2 (for comparison the JET port transmits 46.2 MW m−2 ). Choosing a conservative option of beamline A, the required CD power could be delivered through 16 ports with 4 injectors per port, a total area of 4 m2 , less than 1% of the surface area of the tokamak. The greatest single benefit is to be gained by improving the neutralization efficiency to 95% through the photoneutralizer (D). However given the challenging nature of this development other routes to improvement are considered. Increasing the current density (C) would enable either the number of ports or the number of injectors per unit to be reduced. However, if the packing density is high (say 6 units or more) the differential between an increase in current density (C) and an increase in transmission (B) is reduced. The presence of beamline components, such as calorimeters can have a deleterious influence on the beam transmission. For example, in beamline (A) most of the transmission losses occur in the neutralizer and residual ion dump channels, so there is obviously scope for improvements in design of the beamline. This may be easier to achieve than an increase in current density, thus favouring a high packing density. Conversely, selecting fewer, larger sources (say 4 per beamline) forces development along the path of increased current density as the benefits of increased transmission are depleted. Although increasing the source current density reduces the number of units required, it is clear from Section 3.4 that any such increase should be accompanied by an improvement in the transmission or, more specifically, a reduction in direct interception on beamline components. This is desirable even without an increase in current density if the multi-hour, long pulse mode is to be realised. There is, however, advantage to reducing the number of injectors per unit as it allows greater freedom in determining HV stand-off distances between injectors. To compensate the injector size could increase. Large area sources also yield a higher fraction of extraction area per unit footprint area, when the HV standoff distance is taken into account and in this respect sources with an almost square profile are more efficient than rectangular. One disadvantage of large sources is the degree of steering of the beams into the port aperture that can be achieved. On JET, for instance, the individual injector units are physically oriented at angles of 158 mrad and 51 mrad to the horizontal plane and 61 mrad to the vertical plane in addition to grid inclination and offset aperture steering. On ITER the beam steering is limited to 16 mrad through grid and offset aperture and this contributes to the relatively large port size. Offset aperture steering depends on the magnitude of the electric field in the accelerator and is limited by the onset of beam aberration and direct interception on the grid. It is unlikely that steering greater than that adopted for ITER would be acceptable. It is clear that the only significant improvement is through a combination of all three strategies (E), i.e. increasing current density, improving transmission and increasing the neutralization efficiency. This represents a significant R&D effort over the next two decades. 3.2. Electrical efficiency
Fig. 1. Number of ports required to inject a total of 240 MW for different beamline technologies in Table 1.
Again, the steady state waveform is the critical case; with the heating waveforms the beams are only required for a small fraction of the plasma cycle and so their impact on overall efficiency is reduced. The electrical efficiency for each beamline combination is given in Table 1; the efficiency is estimated by including efficiencies for power supplies and support services such as cryoplant. As before, the single most influential improvement to be made is in the neutralization efficiency. Even if the photoneutralizer were to
E. Surrey et al. / Fusion Engineering and Design 86 (2011) 793–796
795
Fig. 2. Relationship between power per beamline and pulse length for existing fusion experiments and the future JET EP2, ITER and DEMO devices. Note uncertainty on DEMO power indicated by error bars.
prove possible, achieving the desired electrical efficiency of 60% still requires optimistic and simultaneous scaling of the other factors. This is a key point for the deployment on NBI on DEMO: increasing the NBI efficiency from 27% to 60% reduces the installed NB power from ∼890 MW to ∼400 MW (to deliver 240 MW). At 27% NBI efficiency the power output from a 2.5 GW (thermal) reactor with generating efficiency of 40% is almost entirely consumed by the NBI system, necessitating a much larger reactor. 3.3. Caesium consumption The ITER source uses caesium to enhance the surface production of negative ions to produce the required density. The PINI-sized prototype source consumes 14 g of caesium per second of plasma time [9]. It is not clear at present if the caesium consumption is dominated by diffusion out of the source or by gettering at the wall but as the extraction area and source wall area are closely linked in volume sources (at least to date) either mechanism can be used to scale for a DEMO requirement of ∼24 g caesium per 10 h pulse or ∼17 kg per year. Apart from the obvious safety issue of storing such quantities of caesium, some of the metal will diffuse into the accelerator and cause breakdown of insulators. It is clear that an alternative to caesium would be highly desirable and this may require adopting alternative source technology such as the pulsed afterglow plasma [12] and this may require a change in the philosophy of neutral beam injection. 3.4. Long pulse operation 3.4.1. Thermal management The anticipated pulse length of the steady state waveform, upwards of 10 h, is another significant departure from present experience with NB systems. The extent to which DEMO stretches the operational envelope is illustrated in Fig. 2 where the NB power per beamline (or port) is plotted as a function of the pulse length for a number of fusion experiments. The inverse nature of the power transmitted through the beamline and the pulse length in most current experiments is clear and arises from the engineering design of the beamline and its components. The location of the ITER operating space and possible DEMO parameters lie significantly outside current experience and will provide particular challenges in engineering design and material science. Of concern are the many un-cooled components, particularly at the periphery of the beamline, where the incident heat flux is generally regarded as trivial and so are not subject to full design scrutiny. To investigate the effect of low heat flux over a prolonged period a finite element model of a generic component, (a CuCrZr
Fig. 3. Finite element model of generic peripheral beamline component used in long pulse study.
plate bolted to a steel vessel with an average contact pressure of 2.7 M N m−2 , giving a thermal conductivity of 700 W m−2 K−1 for a surface finish of 1–1.5 m [13]), typical of those used for protection on current fusion devices, was constructed. In the model there is no active cooling in the region, and the only available mechanism of external cooling is through natural air convection at the top face of the vessel (i.e. radiation is ignored). The equilibrium temperature distribution (reached in ∼20 h), calculated for a relatively low value steady state incident heat flux of 5 kW m−2 , such as might be caused by interception of stray beam ions, is illustrated in Fig. 3 where a peak temperature of 1200 K is obtained for the most severe conditions considered. Increasing the conductivity at the interface to infinity only reduces the maximum temperature by 6% but a tenfold increase of the convection coefficient at the outer vessel wall reduces the peak temperature by one third. Regarding radiation, we consider the generic component to be enclosed by a surface of finite emissivity and area. Thus the enclosure can also radiate heat back to the component, a process that is characterized by the geometric viewing factor, emissivities, areas and temperatures of the component and the enclosure. The effect of varying the emissivity of either the component or the enclosure and of the viewing factor is shown in Fig. 4 for the conditions of Fig. 3 and equal radiative areas of component and enclosure. There is no preferential effect between the emissivities; higher values lead to lower component temperature, although the effect diminishes above 0.5. Similarly the viewing factor has the largest impact between 0 and 0.5. (The temperature of the enclosure does not have a significant impact below 450 K.) Thus by manipulating the emissivity of the component and enclosure surface and by geometric consideration of the viewing factor, thermal management of peripheral components should be feasible without resorting to active cooling. 3.4.2. Mechanical failure In addition to the thermal aspects of long pulse operation discussed in Section 3.4.1, there are two mechanical failure mechanisms associated with long pulse operation: creep-fatigue and irradiation creep. Creep-fatigue is a combination of normal creep mechanism encountered in continuous operation and fatigue encountered in low duty cycle operation. The NBI system will be subject to cyclic stress superposed on a constant load as a result of either modula-
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emissivity or view factor Fig. 4. Effect on component temperature of variation in emissivity, ε and view factor: component varies enclosure ε = 0.5, enclosure varies, component ε = 0.5, enclosure varies, component ε = 0.25 and solid line view factor varies (both ε = 0.5).
tion in the steady state DEMO or the cyclic waveform for the heating phase of the pulsed DEMO. The combination lowers the lifetime of components significantly; a reduction of almost 50% was recorded by Wu et al. [14] and the effect is temperature dependent [15]. Irradiation creep arises from displacement damage due to neutrons and the production of transmutant helium. The effect is sensitive to temperature and neutron energy and has been studied extensively for the fast fission neutron spectrum but not for the fusion spectrum. A definitive description of irradiation creep combined with creep-fatigue is given in [16], where an example for Austenitic 316 steel shows a reduction in lifetime of an order of magnitude compared to thermal creep only. Current design codes do not address irradiation, although the methodology of [16] could be used if more fusion relevant data were available. 4. Conclusions The possible operational demands on NBI of a DEMO device have been assessed in terms of NB power and pulse length. Both are outside present operational experience and represent significant challenges to the technology. Three key parameters are identified for future development: beam current density, neutralization effi-
ciency and beam transmission and it is shown that improvements in all three must be simultaneously achieved to obtain an electrical efficiency relevant to power generating plant. The requirement for NBI power modulation precludes using a small number of large injectors although this approach is more efficient in terms of spatial use and in this respect square profile sources make the most efficient use of the stacking footprint area. Considering the deployment of the NBI system on the tokamak, there is little point in striving for high packing density per port as the interruption to the blanket is less than 1% of total area. Finally, consideration has been given to the effect of low heat flux acting over prolonged periods on uncooled components, such as may be found in the periphery of the beamline. It is shown that by manipulating the emissivity of the component and enclosure surfaces and by implementing the correct geometry with respect to their enclosure, the temperature of these items can be controlled through radiation cooling alone. Acknowledgements This work was funded by the United Kingdom Engineering and Physical Sciences Research Council under grant EP/G003955 and the European Communities under the contract of Association between EURATOM and CCFE. The views and opinions expressed herein do not necessarily reflect those of the European Commission. References [1] D. Maisonnier, et al., Nucl. Fus. 47 (2007) 1524. [2] D.J. Ward, W.E. Han, Results of System Studies for DEMO, TW6-TRP-002 CCFE, 2007. [3] G.M. Voss, et al., Fus. Eng. Des. 63–64 (2002) 65. [4] F. Najmabadi, et al., Fus. Eng. Des. 80 (2006) 3. [5] D.J. Campbell, et al., IAEA-CN-FT/1-1, in: 21st Fus. Energy Conf., Chengdu, China, 2006. [6] H.P.L. de Esch, et al., Fus. Eng. Des. 26 (1995) 589. [7] T. Suzuki, et al., Nucl. Fus. 48 (2008) 045002. [8] ITER DDD 5.3, N53 DDD 29 01-07-03 R0.1 IAEA, Vienna (2003). [9] W. Kraus, et al., 1st Symp. Neg. Ions, Beam & Sources Aix-en-Provence 2008, in: AIP Conf. Proc., vol. 1097, 2009, p. 275. [10] M. Kovari, B. Crowley, Fus. Eng. Des. 85 (2010) 745. [11] G. Duesing, et al., Fus. Technol. 11 (1987) 163. [12] M.B. Hopkins, et al., J. Appl. Phys. 70 (2009) 1991. [13] W. Rohsenhow, Handbook of Heat Transfer Fundamentals, McGray Hill, 1973. [14] X. Wu, et al., J. Nucl. Mater. 367–370 (2007) 984. [15] S.D. Preston, et al., Fus. Eng. Des. 48 (2003) 527. [16] P.J. Karditsas, Fus. Eng. Des. 48 (2000) 527.