R&D FOR RELIABLE DISRUPTION MITIGATION IN ITER

Transcription

1 M. Lehnen et al. R&D FOR RELIABLE DISRUPTION MITIGATION IN ITER M. Lehnen, D.J. Campbell, D. Hu, U. Kruezi, T.C. Luce, S. Maruyama, J.A. Snipes, R. Sweeney ITER Organization, Route de Vinon sur Verdon, CS , Saint Paul-lez-Durance, France N.W. Eidietis General Atomics, PO Box 85608, San Diego, CA , USA A. Matsuyama National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori , Japan E. Nardon CEA, IRFM, F Saint-Paul-lez-Durance, France Abstract The disruption mitigation system (DMS) is a key plant system to ensure the reliable and successful operation of ITER from the first experimental campaign onwards. The DMS baseline concept and design is based on present knowledge on disruption mitigation, which, nevertheless, remains subject to significant gaps in understanding, especially as concerns runaway electron (RE) formation and mitigation. The paper outlines the challenges of implementing a highly reliable DMS for ITER, presents recent progress towards the consolidation of the baseline system and develops a strategy and plan for achieving the required level of disruption mitigation to satisfy ITER s operational needs. 1. INTRODUCTION Unmitigated disruptions in ITER can lead to melt damage at plasma currents and thermal energies well below the targets of 15 MA and 350 MJ (M. Lehnen et al., 2016). The Disruption Mitigation System (DMS) has to ensure that the thermal energy is dissipated so that the resulting thermal fluxes to the first wall are below melt limits. The following current quench has to be fast enough to prevent large halo currents, but slow enough to avoid excessive eddy current loads on in-vessel components. The DMS, while dissipating the thermal energy, has also to ensure that the formation of runaway electrons (RE) is impossible. High uncertainties remain on the critical energy and current of a runaway electron beam above which melting would occur, but it is evident that the impact of a fully developed runaway beam in a 15 MA disruption could cause cooling water channels to fail. The presently envisaged mitigation principle for ITER relies on energy dissipation by line radiation. The currently foreseen method to generate sufficient line radiation is based on the massive injection of impurities such as neon and argon but also large quantities of deuterium. These will be delivered to the plasma in the form of fragmented cryogenic pellets using the Shattered Pellet Injection (SPI) technology (L. Baylor et al., 2018). This technology is favourable due to its higher delivery rate compared to Massive Gas Injection (MGI). Simulations with the code ASTRA have shown that only up to 10% of the total gas from MGI will enter the plasma before the Thermal Quench (TQ) onset when installed with a flight tube length of 8 m. The efficiency of MGI is thus too low to ensure reliable thermal load mitigation together with RE avoidance. The particle delivery from SPI will be much more compact and therefore more efficient since the pellet is fragmented only at the end of the flight tube just before entering the plasma. Another advantage of SPI is the possible deep penetration of ice fragments into the plasma leading to high assimilation in the plasma core (see discussion in section 4.2). The delay times from DMS trigger to the mitigated thermal quench will be longer for SPI compared to MGI since pellet velocities will be below the sound speed. The delay times for SPI at projected velocities of the order of 150 m/s would be about 30 ms not taking into account the possible delay caused by breaking the pellet loose and by the acceleration by the propulsion gas. This will need to be taken into account when designing the disruption prediction system to trigger the DMS. Successful prediction triggers the DMS such that the injection takes place before the thermal quench would occur. To reduce the heat loads and to avoid runaway electrons it is foreseen to inject a combination of neon, to radiate the thermal energy, and deuterium to increase the plasma density to values that prevent the formation of runaway electrons (J.R. Martín-Solís et al., 2017). This injection will consist of mainly pellet fragments of adequate size to ensure deep deposition. Should prediction be unsuccessful, it will be essential to detect an ongoing disruption to still reduce heat loads and electro-magnetic forces during the current quench. This is essential since it was found with the ITER-like wall in JET that most of the magnetic energy is dissipated through the halo when 1

2 IAEA-CN-258/318 the plasma is vertically displaced in the current quench. This conversion of magnetic energy into thermal energy can lead to significant melting of the beryllium first wall and make a high mitigation success rate a necessity during high current operation (M. Lehnen et al., 2016). The post-tq injection has to have a high fraction of gas to ensure sufficient assimilation in the colder current quench plasma. As a second layer of protection, argon injection is envisaged to enhance the radiative dissipation of the energy of a runaway beam should it form accidently (see discussion in section 4.5). The injection schemes are shown in FIG. 1. FIG. 1. Envisaged mitigation schemes. Left: Thermal quench mitigation with neon injection and runaway avoidance through massive injection of deuterium.right: Should a runaway beam form argon injection is foreseen to dissipate the energy. 2. PRESENT LAYOUT OF THE DISRUPTION MITIGATION SYSTEM In view of the expected injection quantities required for the foreseen mitigation schemes, it is presently considered to extend the space allocation for the DMS to allow the injection of up to 32 pellets from three equatorial port plugs (for thermal quench mitigation, runaway avoidance, and runaway energy dissipation) while utilising three upper port plugs for post-tq mitigation of current quench loads (see FIG. 2). A preliminary layout of eight injectors inside an equatorial port plug is given in FIG. 3. The diameter of each pellet in the equatorial port plugs will be 28.5 mm with a length over diameter ratio of L/D = 2. This allows injecting up to 0.9x10 24 Ar atoms or 1.1x10 24 D atoms with each pellet. Presently, RE avoidance and RE energy dissipation are FIG. 2. Presently proposed DMS configuration in the equatorial plane (L1) and for the upper ports (L2). Three equatorial port plugs will be equipped with a total of up to 32 injectors (eight in each for EP#08 and EP#17 and sixteen in EP#2). Provisions will be made to allow a possible reconfiguration at a later stage in the project if required (EP #11).The configuration in L2 (upper port plugs) remains as previously planned but with single pellet injection from each expected to require of the order of of the port plugs. particles of each species. However, uncertainties on these quantities are high (see discussion in sections 4.4 and 4.5). Considering these numbers, a total of 10 pellets would be required for RE avoidance and 12 pellets for RE energy dissipation when approaching the baseline plasma scenario at 15 MA. Thermal and current quench mitigation on the other hand will require only a relative low quantity of neon of 5x10 22 at maximum. This quantity would be admixed to the deuterium pellets to be injected prior to the thermal quench. The radiation flash and the associated heat fluxes to the first wall components can be critical and injecting

3 M. Lehnen et al. from several locations is a way to compensate for this (see discussion in section 4.1). Additionally, smaller pellet sizes will be foreseen in the upper port plugs for post-tq injection as described in the previous section. FIG. 3. Preliminary layout of eight injectors in one third of an equatorial port plug. The installation includes (from right to left) the propellant gas valve, the freezing cell, a microwave cavity, the primary vacum gate valve, and a bellows at the closure plate of the port plug for mechanical decoupling. At the end of the flight tube inside the port plug, a bend will ensure breaking the pellet into fragments for better assimilation inside the plasma. FIG. 4. Timeline for the R&D on the baseline DMS and the possible upgrades. Decision points for the baseline system are given in red. First plasma (FP) is scheduled for end of 2025 followed by three experimental campaigns (Pre-Fusion Plasma Operation PFPO and Fusion Plasma Operation FPO). 3. OVERALL R&D STRATEGY AND TIMELINE The ITER R&D strategy involves two branches, one concentrating on establishing the best basis for the baseline system that will be implemented before the first experimental campaign, the other assessing the performance of the baseline concept and at the same time exploring and developing alternative schemes. The design of the baseline DMS must be finished in the next couple of years, whereas later upgrades of the system allows R&D on the feasibility of other injection techniques or completely new mitigation strategies on a longer timescale. An overview on the timeline is given in FIG. 4. The R&D on the baseline DMS focusses on four main tasks. The most critical of which is to assess and confirm the feasibility of effectively injecting multiple pellets, 3

4 IAEA-CN-258/318 which is the basis for achieving the required injection quantities. The feasibility of the technology was shown and it is applied now at DIII-D and JET. However, optimisation is required to fulfil the challenging requirements for the ITER system. R&D on the flight tube front end is related to deciding on the right fragment sizes and how to reliably produce them. Finally, the ITER research plan foresees a significant amount of time for commissioning and testing the DMS. However, a basis for this must be established beforehand to ensure efficient procedures that minimise the risk of damaging plasma facing components. Should an upgrade of the DMS be required, there are presently two possible scenarios. If present tokamak experiments and modelling activities show deficiencies of the concept and if an alternative has been developed, design work could start around the time of the first plasma such that the upgrade is available for the assembly phase after PFPO-1 (PFPO Pre-Fusion Power Operation). The risk that the present DMS would not be effective enough to achieve the targets in PFPO-1 is low, since operation will be restricted to well below 15 MA with low thermal energies. In the second scenario, the deficiencies would be detected only during PFPO-1. This would impose a higher risk on achieving the goals of PFPO-2, since installation of the upgrade would only be possible after that campaign. Most important for any upgrade scenario is the availability of alternative technology or mitigation schemes and developing these plays an essential role in the overall strategy. 4. MAIN CHALLENGES STATUS AND R&D NEEDS 4.1. Multiple Injection The injection quantities that are required based on present knowledge can only be achieved by multiple pellet injection. Any bigger pellet sizes to decrease the number of pellets would require long pellet preparation times that would not be compatible with the planned pulse repetition rates. Initial experiments on multiple injections have been done at DIII-D using two injectors toroidally separated by 120 degrees (N.W. Eidietis et al., 2018). These experiments have shown that the relative pellet arrival time at the plasma edge is critical. It was observed that the radiated energy varies significantly when the relative arrival time is varied over about 1-2 ms. It could be expected that the duration from pellet arrival to thermal quench initiation (pre-tq duration) for a single pellet is the critical time window in which both pellets must arrive to be effective. However, this duration could also be reduced when injecting multiple pellets from different locations. More experiments are planned at DIII-D to study such effects. With the thermal energies in ITER it can be expected that timescales are longer and that fragments will subsequently move the cold front towards the plasma center before the onset of the thermal quench. For example, the penetration depth at which a 2 mm fragment will be fully ablated is about 3 cm for the ITER baseline scenario of 350 MJ of plasma energy. Thus, velocity dispersion, pellet arrival times, fragment sizes and the position of the q=2 surface will play equally important roles for the efficiency of multiple injection. More tokamak experiments with at least two injectors in the same location and in the ITER-like toroidal distribution are required taking into account that the penetration depth of single fragments should be much less than the minor radius of the plasma. Studies of ablation processes, impact of velocity dispersion, and/or pellet timing can be done by implementing appropriate ablation and impurity radiation models in 1D or 2D fluid codes. On the other hand, 3D MHD modelling will be required to understand the impact of MHD instabilities, especially on the onset of the thermal quench. However, these codes will have simplified impurity and pellet models. Initial single pellet injection simulations for JET and ITER have been done (D. Hu et al., 2018) (C. Kim et al., 2017) and will be further refined along with the validation of the physics model by comparison to experimental results. Another aspect that is expected to require injection from different locations around the torus is the heat load through the radiation flash at the thermal quench. Experiments with MGI have shown that both injection geometry and MHD play a role in determining the peaking of the radiative heat flux on the first wall and peaking factors in the toroidal direction of below 2 were reported from DIII-D (D. Shiraki et al., 2015) and JET (M. Lehnen et al., 2015). Poloidal peaking was found to range from 1.6 to 2.2 for MGI in DIII-D (N.W. Eidietis et al., 2017). Assuming a peaking of 2 in poloidal and toroidal direction in ITER, a simple estimate shows that above 70 MJ of thermal energy the stainless steel of the diagnostic first wall can show surface melting (M. Lehnen et al., 2015). The progressive surface roughening would determine how many mitigated disruptions can be tolerated at specific energies, constraining the allowable disruption rates in the various phases of operation. With SPI, a more localised and radially deeper deposition of the radiating impurity species in the toroidal and poloidal directions is expected. In how far this will impact radiation peaking considering the deeper deposition of material inside the plasma and

5 M. Lehnen et al. by multiple injection locations has to be studied in experiment and modelling. Initial modelling of the planned SPI in JET with the JOREK code shows strong poloidal and toroidal peaking despite an artificially broad impurity distribution in the toroidal direction for numerical reasons (D. Hu et al., 2018) Industrialization of the SPI technology The ITER environment and the challenging requirements for the DMS will require further design development to demonstrate readiness of the technology and its peripheral systems for ITER. Reliability and reproducibility will require improvements in the understanding of the pellet formation process and the pellet shearoff & acceleration. Input on the requirements for these processes will partly come from tokamak experiments as described in the other sections. Space restrictions and the need for openings of the flight tube for diagnostics and the recovery of the propellant gas will require tests on critical impact angles to ensure the pellet is not breaking inside the flight tube. For commissioning and also for reliably operating the DMS, diagnostics have to be developed to monitor the integrity of the pellets and to optimise pellet synchronisation. The possible accumulation of gases used during an ITER pulse on the pellets and the impact of such accumulation on the pellet integrity and the mitigation performance needs to be studied. Furthermore the impact of propellant gas entering the plasma together with the pellet fragments as well as the consequences of prematurely broken pellets especially in view of runaway electron generation have yet to be studied in experiments and modelling. These types of studies are necessary to quantify risks and to define the requirements for the injector more precisely Requirements on fragment sizes and velocity Fragment size and velocity as well as its dispersion are key parameters that will determine the dynamics of the cooling down process before the thermal quench onset and therefore the assimilation efficiency. These parameters will have to be chosen such that efficient mitigation is ensured for all target plasmas which will have a large range of currents and energies. Especially when close to a disruption, the discharge is likely to have developed instabilities that can lead to significantly different dynamics during the pre-tq phase. For quantification of the impurity deposition it will be required to have a good understanding of the pellet ablation process using local models that are precise in atomic and ablation physics and to make use of global models that take into account transport processes but with approximations in the impurity source description (c.f. (D. Hu et al., 2018) and (C. Kim et al., 2017)). Pellet ablation models exist for hydrogenic species (P.B. Parks and R.J. Turnbull, 1978) and were extended to impurity species like neon (P.B. Parks, 2017) and towards descriptions of multiple fragment injection (R. Samulyak et al., 2018). Experimental validation for neon is still pending and would require well defined experiments, likely with single, small impurity pellets instead of the complex injection of many fragments of different sizes. SPI experiments on tokamaks of different size and thermal energies with flight tubes and shattering techniques specifically designed to explore the efficiency of different fragment size distributions are required. This will likely also require development work in laboratories to develop these techniques. Here, input from modelling and theory activities is needed for guidance on fragment sizes, gas fractions and fragment velocities. Especially the sensitivity on assimilation and core density rise and their dependency on these parameters needs to be clarified. For any SPI experiment, it will be essential to know the characteristics of the fragments for different injection parameters like pellet composition and velocity investigated in laboratory tests Runaway Avoidance The highest priority when activating the DMS is to avoid the formation of runaway electrons. Presently, this is foreseen by the injection of deuterium (or hydrogen) before the thermal quench. This scheme prevents formation of runaways in JET (M. Lehnen et al., 2011), but cannot be simply scaled to ITER. Simplified models predict that quantities on the order of deuterium atoms are required in ITER (J.R. Martín-Solís et al., 2017) (P.B. Aleynikov, 2018). More complex modelling using 3D MHD codes and appropriate runaway models within these codes are required to fully understand the role of MHD on this process (like for example in (A. Matsuyama et al., 2017)). Test particle studies with NIMROD and JOREK have shown that field line stochasticity during the thermal quench is not sufficient to ensure the entire loss of seed electrons (V.A. Izzo et al., 2011) (C. Sommariva et al., 2018). Moreover, the density rise inside the core plasma can be expected to be non-uniform as shown in 5

6 IAEA-CN-258/318 FIG. 5. This has to be taken into account when assessing the efficiency of runaway avoidance during the thermal quench. On the experimental side, the diagnosis of the thermal quench dynamics and the detection of small runaway seed quantities are most challenging. But reliable extrapolation from SPI experiments at relatively low current compared to ITER is impossible without the use of validated models, since the critical seed population sufficient to generate a substantial runaway beam decreases from about 10 8 (JET at 5MA) to ~10 3 in ITER at 15 MA. Note that the highly conductive vacuum vessel is decreasing the avalanche multiplication in ITER compared to JET with a highly resistive vessel (M. Lehnen et al., 2017). Recently, the impact of instabilities (MHD and kinetic) on runaway formation during disruptions has been reported in experiments (e.g. (L. Zeng et al., 2013) and (A. Lvovskiy et al., 2018)). If there is no alternative technique to avoid runaway formation, these instabilities can play a relevant role in the overall process and further investigations in theory and experiment are required. It is important to mention here that diagnostics on present machines are not optimised for providing answers on disruption mitigation issues. Diagnostic development and implementation are needed to answer questions on radiation distributions and to provide absolute values for the radiated energy to quantify the effectiveness of the injection (R. Sweeney et al., 2018). Similarly, density measurements during disruptions are line integrated and no spatial information is presently available, but urgently needed to better understand RE avoidance. FIG. 5. JOREK SPI simulations for a JET plasma showing the non-uniform distribution of free electron density towards the end of the thermal quench Runaway Energy Dissipation In case of an accidental runaway formation, a second layer of protection is desired to dissipate the kinetic and magnetic energy of the runaway beam. The present baseline approach foresees the injection of argon into the current quench phase pre-emptively to radiate the energy of the runaways, should they form. The required dissipation rate depends on the vertical displacement of the runaway beam that will follow the drop of current from the initial to the runaway plateau level 1. Initial simulations using the code DINA (S. Konovalov et al., 2016) show that not only the displacement time is relevant (on the order of 100 ms), but that also the scraping-off of the runaway beam while it is moving into the first wall might require significantly higher assimilated argon quantities than expected from kinetic models (P. Aleynikov et al., 2014) (L. Hesslow et al., 2017), of the order of argon atoms instead of 5x The ITPA MHD group is presently assessing the dependence of di RE/dt during the plateau phase as a function of the injected quantity of high-z impurities. This characterizes the dissipation of magnetic energy, which will be converted into runaway kinetic energy, if not dissipated through radiation. Note, that the expected kinetic energy of a worst case 10 MA runaway beam in ITER is about 20 MJ whereas its magnetic energy is expected to be up to 300 MJ (J.R. Martín-Solís et al., 2014). Assuming that during the plateau phase the electric field is close to the critical electric field, E E c, the current decay can be written as 1 Position control is only possible with provisions if the drop in current is less than about 30% (V.E. Lukash et al., 2013).

7 M. Lehnen et al. di RE dt = 2π (l + l 1 i μ 0 2 ) E c, and is thus proportional to the impurity density in the plasma with which the runaways interact. The part of the inductance external to the vacuum vessel can be neglected if the current decay is faster than the vessel resistive time, thus for ITER l = log(8r a) 2 0 (R = major radius, a = minor radius). A decrease of runaway current is observed in many devices after the injection of high-z impurities. However, the reported assimilation efficiencies are low (G. Pautasso et al., 2017) (Hollmann, 2013). A peculiarity is observed in JET for which an acceleration of the current decay is only observed if the initial impurity injection to generate runaways is small (C. Reux et al., 2017). First injections with SPI in DIII-D show similar performance as MGI (D. Shiraki et al., 2018). Additionally, these experiments show that deuterium injection (as foreseen in ITER for runaway avoidance) can have a negative impact on the energy dissipation rate. The background plasma temperature plays a crucial role for impurity assimilation. For example, the plasma is transparent to neutral argon at T e < 2eV. At higher T e the penetration timescale is determined by transport processes and is likely to be too long to achieve the required energy dissipation rates in ITER. The temperature of the background plasma in turn depends on the energy balance between the power input from runaway-thermal collisions and the losses through line radiation, which is dominating the loss process (E.M. Hollmann et al., 2015). Even if impurities could penetrate into a cold plasma, their interaction with the runaways will immediately increase the temperature. This effect could, for example, explain the findings at JET. These issues have to be addressed to possibly identify solutions or to document the performance that can be expected for this scheme in ITER. Further understanding of these processes will require quantifying accurately the parameters of the background plasma. Another aspect to be studied is the MHD instabilities, which could cause deposition of runaways on the wall components before their energy is reduced to an acceptable level. 5. SUMMARY AND CONCLUSIONS The DMS is an essential plant system to ensure that ITER can achieve its goals. However, the technique on which the DMS is based, SPI, has been explored so far at only one tokamak. Recently, J-TEXT has been equipped with an SPI system and JET will have SPI available in its coming campaign. However, increasing confidence about the capabilities of the ITER system will require a significant increase of R&D in this area and especially more tokamak experiments to allow more reliable scaling towards ITER. The DMS R&D plan foresees dedicated experiments targeted to address some of the most urgent questions as outlined in this paper. A new international Task Force has been formed under the auspices of the ITER Organization to implement the R&D plan. It will work goal-oriented along specific R&D tasks to ensure a better basis for the baseline DMS. Besides the urgent need to optimise the baseline approach, it is of equal importance to enhance R&D on the exploration of alternative techniques, either for the injection technology or for the mitigation scheme itself. Here, the efforts in the domestic programmes of the ITER members will be essential to find new approaches and to prove a maturity level that allows considering them for ITER. DISCLAIMER ITER is the Nuclear Facility INB no The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. This publication is provided for scientific purposes only. Its contents should not be considered as commitments from the ITER Organization as a nuclear operator in the frame of the licensing process. 7

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