# Disruption avoidance strategies for DEMO

Wolfgang Biel, E. Alessi, R. Ambrosino, Marco Ariola, Inessa Bolshakova, Kai Jakob Brunner, Marco Cecconello, Sean Conroy, Dejan Dezman, Ivan Duran, et al.

► To cite this version:

Wolfgang Biel, E. Alessi, R. Ambrosino, Marco Ariola, Inessa Bolshakova, et al. Disruption avoidance strategies for DEMO. 28th European Fusion Programme Workshop, Dec 2020, visioconference, France. hal-03096054

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EUROfusion

# Disruption avoidance strategies for DEMO

EFPW

02nd – 04th Dec 2020

## Wolfgang Biel¹,² and contributors to EUROfusion WPDC

¹Institute of Energy- and Climate Research, Forschungszentrum Jülich GmbH, Germany

²Department of Applied Physics, Ghent University, Belgium

w.biel@fz-juelich.de

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JÜLICH
FORSCHUNGSZENTRUM

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GHENT
UNIVERSITY

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This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
W. Biel¹,², E. Alessi¹⁰, R. Ambrosino³, M. Ariola⁴, I. Bolshakova¹⁷, K. J. Brunner¹⁶, M. Cecconello⁵, S. Conroy⁵, D. Dezman¹³, I. Duran⁶, S. Entler⁶, E. Fable⁷, D. Farina¹⁰, Th. Franke⁷,⁹, L. Giacomelli¹⁰, L. Giannone⁷, R. Gomes¹¹, B. Gonçalves¹¹, S. Heuraux¹⁵, A. Hjalmarsson⁵, M. Hron⁶, F. Janky⁷, A. Jesenko¹³, A. Krimmer¹, O. Kudlacek⁷, R. Luís¹¹, O. Marchuk¹, G. Marchiori⁸, M. Mattei¹⁸, F. Maviglia²⁰,⁹, G. De Masi⁸, D. Mazon¹², P. Muscente¹⁰, Y. Nietiadi¹¹, S. Nowak¹⁰, A. Pironti³, A. Quercia³, E. Ricardo¹¹, N. Rispoli¹⁰, G. Sergienko¹, R. Schramm⁷, S. El Shawish¹⁴, M. Siccinio⁷,⁹, A. Silva¹¹, F. da Silva¹¹, C. Sozzi¹⁰, M. Tardocchi¹⁰, D. Testa¹⁹, W. Treutterer⁷, A. Vale¹¹, O. Vasyliev¹⁷, S. Wiesen¹, H. Zohm⁷

¹Institut für Energie und Klimaforschung, Forschungszentrum Jülich GmbH, Germany

²Department of Applied Physics, Ghent University, Belgium

³Università degli Studi di Napoli Federico II, Consorzio CREATE, Italy

⁴Università di Napoli “Parthenope”, Consorzio CREATE, Italy

⁵Department of Physics and Astronomy, Uppsala University, Sweden

⁶Institute of Plasma Physics, Czech Academy of Science, Praha, Czech Republic

⁷Max-Planck-Institut für Plasmaphysik, Garching, Germany

⁸Consorzio RFX (CNR, ENEA, INFN, Università di Padova), Padova, Italy

⁹EUROfusion Power Plant Physics and Technology (PPPT) department, Garching, Germany

¹⁰ISTP-CNR-Mi, Istituto di Scienza e Tecnologia dei Plasmi, Milano, Italy

¹¹Instituto de Plasmas e Fusão Nuclear, IST, Universidade de Lisboa, Portugal

¹²CEA, IRFM F-13108 Saint Paul-lez-Durance, France

¹³Cosylab, Ljubljana, Slovenia

¹⁴Jožef Stefan Institute, Ljubljana, Slovenia

¹⁵Institut Jean Lamour UMR 7198 CNRS-Université de Lorraine, ARTEM BP 50840, F-54011 Nancy, France

¹⁶Max-Planck-Institut für Plasmaphysik, Greifswald, Germany

¹⁷Magnetic sensor laboratory, Lviv, Ukraine

¹⁸Università della Campania “Luigi Vanvitelli”, Consorzio CREATE, Italy

¹⁹Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland

²⁰Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, C.P. 65-00044 Frascati, Rome, Italy

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JÜLICH
FORSCHUNGSZENTRUM

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GHENT
UNIVERSITY

ENEA

Agencia nazionale per le nuove tecnologie,
l'energia e lo sviluppo economico sostenibile

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UPPSALA
UNIVERSITET

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INSTITUTE OF PLASMA PHYSICS
OF THE CZECH ACADEMY OF SCIENCES

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INSTITUTO DE PLASMAS
E FUSÃO NUCLEAR

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Jožef Stefan
Institute

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MSL

EPFL

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COSYLAB

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JÜLICH
FORSCHUNGSZENTRUM

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GHENT
UNIVERSITY
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* Pulsed DEMO tokamak as baseline

* Machine size ~ 1.5 times of ITER

* Conservative plasma scenario with

* predominantly ohmic current drive

* confinement quality $H_{H98y2} \sim 1$

* high core radiation fraction but $P_{heat}-P_{rad,core} > P_{LH}$

* high plasma density with fully detached divertor plasma operation (lower single null divertor)

* no ELMs or very small ELMs (factor 100!)

* Vertical maintenance concept for blanket

* only limited access for diagnostics from the top side

* Large equatorial ports

* main access for diagnostics

<figure><img src="image_22.png" /></figure>

## Main DEMO parameters:

* $R_0$, a ~ 9 m, 2.9 m

* $P_{therm}$ ~ 2 GW

* $P_{el,net}$ ~ 300...500 MW

* $t_{pulse}$ > 2 h

* $W_{kin}$ ~ 1.3 GJ

* $V_{plasma}$ ~ 2500 m³

* $A_{wall}$ ~ 2000 m²
<figure><img src="image_23.png" /></figure>

## Heat impact factor:

$$η = W_{th}/(A_{eff}\sqrt{t})$$

## Critical heat impact factor (melting):

$$η_{melt} = (T_{melt} - T_{op})\sqrt{\frac{\pi \lambda \rho c}{4}}$$

where:

$T_{melt}$ = melt temperature

$T_{op}$ = operational temperature

$\lambda$ = heat conductivity (W/mK)

$\rho$ = mass density (kg/m³)

$c$ = heat capacity (J/kgK)

## Material data for tungsten:

$\lambda$ = 170 W/mK

$\rho$ = 19300 kg/m³

$c$ = 138 J/kgK

## Critical limits for tungsten:

$$η_{melt} = 50 \text{ MJ/(m}^2\text{s}^{0.5})$$

$$η_{crack} = 5 \text{ MJ/(m}^2\text{s}^{0.5})$$

## Major (non-mitigated) disruption

* About 50% of the thermal energy (1.3 GJ) is released within the quench time $t_{TQ} = 1...4$ ms

* Most of the energy is deposited in the divertor

* Peak heat loads $\rightarrow \eta > 130 \text{ MJ/(m}^2\text{s}^{0.5})$

* Large area melting of divertor surface with typically several hundred µm depth

* The heat load capability of the divertor can be affected after a single major disruption

## Mitigated disruption

* About 60% of the thermal energy (1.3 GJ) is released within the quench time $t_{TQ} = 1...3$ ms

* Most of the energy is deposited at the first wall with a typical peaking factor 3

* Peak heat loads $\rightarrow \eta > 30 \text{ MJ/(m}^2\text{s}^{0.5})$

* Near to melting in highly loaded areas, otherwise crack formation

* Several mitigated disruptions may be tolerable if the startup limiters (OML) are not damaged
1) Define the plasma scenario and machine design with sufficient margins against any operational limits

* See DEMO control requirements

2) Ensure redundancy and reliability of diagnostics and actuators for plasma control

* See overview table

3) All elements of the plasma control system (measurements, actuators, controllers) have to be fast and strong enough to cope with any disturbances acting on the plasma

* Plasma dynamics being analysed via control simulations

* Requirements to be defined accordingly

* Model-based and multivariable control for accurate and robust control

* Exception handling for adequate and fast decision taking

* Actuator management to make best use of the resources

4) Make DEMO resilient against those disruptions that cannot be avoided...

* Protection limiters

... and of course provide disruption mitigation schemes (not addressed here)
# 4) DEMO protection limiter concept

## Transients

Limiters are high flux (i.e. operable up to ~10 MW/m²), toroidally and poloidally discontinuous structures which prevent plasma/wall contact during planned and unplanned transients.

* **Upper limiter (UL):**

* U-VDE

* **Inner Midplane limiter (IML):**

* loss of confinement (plasma shrinks)

* **Outer Midplane limiter (OML):**

* Ramp up

* **Outer Lower limiter (OWL):**

* D-VDE

<figure><img src="image_24.png" /></figure>

<figure><img src="image_25.png" /></figure>
<figure><img src="image_26.png" /></figure>

# 1) Control margins and baseline definition

* According to the confinement scaling, the highest fusion power is expected when operating as near as possible to the operational limits (marginally stable solution against infinitesimally small disturbance)

* However, experiments and control simulations show that the plasma reacts with major parameter changes on any small disturbance.

* Reliable operation requires keeping sufficient distance from operational limits (→ control margins)

<figure><img src="image_27.png" /></figure>

A proper compromise has to be found between high fusion power and enhanced stability of operation.
1) Equilibrium control during flat-top phase:

a) The plasma current $I_p$ should be controlled within +/- 0.5 MA, such that the edge safety factor during flat-top operation always remains in the interval $q_{95} = 3.5 \pm 0.1$ (presumably no problem here)

b) The horizontal and vertical position of the current centroid should be controlled such that an initial displacement of 15 cm can always be recovered, where the minimum distance between separatrix and first wall and limiters in all poloidal locations always remains larger than 8 cm, in order to minimize the wall loads.

* Initial DEMO equilibrium control simulations with external PF coils did not meet this requirement.

* We will employ more "advanced" control, but we might need also:

* stronger (faster) external PF coils (this might marginally work), and/or

* additional in vessel control coils (simulations show that this would work), and/or

* larger clearance between plasma and wall, and/or

* reduction of plasma shaping, and/or

* a combination of all these

<figure><img src="image_28.png" /></figure>

c) The plasma shaping has to be controlled such that both the elongation and the triangularity remain within +/- 0.03 of their absolute target values, respectively.

* This new requirement might require a change of the baseline:

* Stable control needed also at the upper limit of elongation → reduction of $K$ needed ?
2) Kinetic control during flat-top phase:

a) The electron density at the separatrix should be controlled at a level of 0.45 $n_{GW}$ (Greenwald limit) with a margin of +/- 0.05 $n_{GW}$ (**presumably no problem**)

b) The fusion power should be controlled to keep the variations of fusion power always below +/- 20 percent over not more than 20 seconds. (??)

c) The power loss $P_{sep}$ across the separatrix should be controlled to maintain a value of 1.2 times the confinement mode threshold (e.g. $P_{LH}$), where deviations of up to additional 50 MW within up to 20 seconds represent the tolerable limit. → **Compatibility to divertor protection needs to be shown.**

d) The divertor plasma should be controlled to keep strongly detached plasma conditions, i.e. a high density plasma with a plasma temperature in front of the target of less than 2 eV. It is assumed that under these conditions both the heat flux density and the erosion rate at the divertor targets are tolerable. Any loss of detachment ($T > 3$ eV) has to be recognized by the control system within less than 0.2 seconds.
→ **dynamic divertor plasma modelling needed for validation.**

a) The plasma pressure should be controlled such that the normalized plasma beta will not exceed 90% of the ideal beta limit (**presumably no problem here**)
### 3) Control of unforeseen events:

a) In case of island formation in the plasma, an island width of $w = 6$ cm should be clearly detectable by the control system within less than 0.1 seconds.

* Looks good for fast rotating islands (> 100 Hz); for slow islands to be investigated

b) The control system should be able to recognize any unplanned increase of plasma radiation greater than $dP_{rad}/dt > 0.1 W_{kin}/T_E^2$ ("impurity events") within less than 0.2 seconds.

* Caveat: too large impurity events can be easily recognized but there might no counter-measures available 😞

<figure><img src="image_29.png" /></figure>

c) The DEMO plasma should be ELM-free, or the ELMs should be smaller than 1 percent of the "natural ELMs" (the latter would deposit a typical energy onto the divertor targets of $E_{ELM,nat} \sim 10$ MJ/m$^2$) – ref. M. Siccinio

* Caveat: reliable control of ELM size by a factor > 100 with not more than a few ELMs per fpy has never been shown → consequently, alternatives to H mode are now being considered

(ELM free plasma scenarios see sessions 1+2)
## 2) Redundancy in the DEMO D&C concept

<figure><img src="image_30.png" /></figure>

<table><thead><tr><th></th><th>Control quantity</th><th>Operational limits</th><th>Diagnostics</th><th>Actuators + interactions</th></tr></thead><tbody><tr><td rowspan="2">equilibr. control</td><td>Plasma current</td><td>safety factor limit (q<sub>95</sub>)</td><td>Magnetic diagnostics<br>Faraday sensors</td><td>CS+PF coils<br>Heating+current drive (HCD)</td></tr><tr><td>Plasma position and shape,<br>incl. vertical stability</td><td>Wall loads (FW and div.)<br>Max. Δz / VDE disruption</td><td>Magnetic diagnostics<br>Reflectometry, ECE<br>Neutron/gamma diagnostics<br>IR polarimetry/interferometry<br>Radiation power profile</td><td>PF+CS coils<br>HCD</td></tr><tr><td rowspan="3">kinetic control</td><td>Plasma edge density</td><td>Density limit<br>Divertor detachment access</td><td>Reflectometry<br>IR polarimetry/interferometry<br>Spectroscopy (pellet injection)</td><td>Pellet injection (fuel)<br>Gas injection</td></tr><tr><td>Fusion power</td><td>wall loads (FW and div.)<br>LH threshold</td><td>Neutron diagnostics<br>FW/blanket and div. power (for<br>calibration only)</td><td>HCD<br>Pellet injection (DT ratio)<br>Impurity gas injection<br>PF+CS coils (shaping)</td></tr><tr><td>Power flow across separatrix</td><td>Radiation limit<br>LH threshold, divertor loads</td><td>Spectroscopy+radiation meas.<br>U<sub>loop</sub></td><td>Impurity gas injection<br>HCD</td></tr><tr><td rowspan="2">instabilities/events</td><td>Divertor detachment and<br>heat flux control</td><td>Divertor wall loads<br>LH threshold</td><td>Spectroscopy+radiation meas.<br>Thermography<br>Divertor thermo-currents</td><td>Gas injection (impurities + fuel)<br>Pellet injection (fuel)<br>PF coils</td></tr><tr><td>Mode control</td><td>various (→ disruptions)</td><td>Reflectometry, ECE<br>IR polarimetry/interferometry<br>Magnetic diagnostics<br>Neutron/gamma diagnostics<br>Radiation power measurements<br>U<sub>loop</sub></td><td>HCD<br>CS+PF coils</td></tr><tr><td></td><td>Unforeseen events<br>(e.g. impurity ingress)</td><td>various (→ disruptions)</td><td>All</td><td>all</td></tr><tr><td></td><td>Disruption mitigation</td><td></td><td>all</td><td>Matter injection</td></tr></tbody></table>
# Risk mitigation via redundancy: Reflectrometry for equilibrium control

<figure><img src="image_31.png" /></figure>

<table><tr><td>Diagn. methods:</td><td><ul><li>reflectometry</li><li>ECE</li></ul></td><td>In case of 2 full poloidal sectors (&lt; 25 cm each) the effect on TBR would be ΔTBR &lt; 0.01</td></tr><tr><td>Control tasks to be covered:</td><td><ul><li>n<sub>e</sub> profile in gradient region</li><li>T<sub>e</sub> profile</li><li>plasma position and shape</li><li>plasma instabilities (up to kHz frequencies)</li></ul></td><td></td></tr><tr><td>Implementation:</td><td><ul><li>&gt; 100 antennae distributed toroidally and poloidally (LFS, HFS, upper, x-point)</li><li>integration of antennae + waveguides in slim poloidal diagnostic cassettes</li><li>routing of waveguides to upper port (attached to „blanket bananas“)</li><li>Microwave feedthroughs or ceramic windows near upper port</li></ul></td><td></td></tr></table>

A. Silva,
A. Malaquias,
C. Sozzi,
G. de Masi, et al.

<figure><img src="image_32.png" /><figcaption>Integration study of MW antennae</figcaption></figure>

## Predicted gap accuracy:

green: high (few cm)

red: low (~ 10 cm)

<figure><img src="image_33.png" /></figure>

R. Luis et al., IEEE 2018

G. Marchiori et al., IEEE 2018
## 3) „Strong control?“: Simulation of equilibrium control using ex-vessel PF coils in a loss of confinement event

<figure><img src="image_34.png" /></figure>

(time traces of $\beta_{pol}$ and $\ell_i$ from E. Fable were used as input)

M. Ariola et al. P 3.045, SOFT 2018

<figure><img src="image_35.png" /></figure>

<figure><img src="image_36.png" /></figure>

<figure><img src="image_37.png" /></figure>

<figure><img src="image_38.png" /></figure>

<figure><img src="image_39.png" /></figure>

→ This control action is too weak; the plasma distance to first wall becomes too small
The required electrical power is nevertheless high
I. Core plasma kinetic control simulation shows strong variation of fusion power (→ tendency to accept some variation, see requirement 2b)

* strong auxiliary heating is the only „fast“ actuator for kinetic control

* requires hundreds of MW electrical power, which would reduce $P_{net,el}$

II. Loss of confinement events lead to inboard movement of plasma

* Current equilibrium control simulations show electrical power demands of > 700 MW without achieving reliable control

* The power requirements follow from Shafranov equation (necessary change of vertical field Bz in a large volume) – would be much lower with in vessel control coils

* Control action may include auxiliary heating in addition (would increase power needs)

III. Vertical control for strongly shaped plasmas

* Current simulations show additional > 100 MW power requirement for ex vessel PF coils, less with in vessel coils

* Electrical power requirement can be reduced by reducing plasma shaping, on the expense of less fusion power

→ provision of sufficient electrical power (~ 1 $GW_{el}$) is a mandatory element of reliable control → electrical storage system needed

<figure><img src="image_40.png" /></figure>
# Tungsten flake (simulation)

* 3 mg of a tungsten penetrating into the plasma

- **FF** vs. **FB**

- FB on $P_{\text{fus}}$ ($P_{\text{NBI}}$), $P_{\text{sep}}$ (Xe)

- „Old“ electric field shear LH model with high hysteresis

* No HL transition occurred

- „New“ $P_{\text{sep}}/P_{\text{LH}}$ model without hysteresis

* **radiation collapse**

* Realistic physics models are extremely important to answer DEMO questions

* NB: 3mg W in plasma is != W falling from the wall (**no model for that**)

<figure><img src="image_41.png" /></figure>
<figure><img src="image_42.png" /></figure>

1) Plasma equilibrium, kinetic and MHD control against „normal disturbances“

→ Margins in the plasma scenario (keep distance from operational limits)

→ reducing the fusion power

→ Margins in the machine design (e.g. limited elongation and shaping)

→ reducing the fusion power

→ Design of diagnostic&control system towards accuracy and performance

→ high effort and cost

2) Failure or control components or control system

→ achieve high reliability standards and install redundancy (number of channels and methods)

→ high effort and cost

3) Human error

→ should be very unlikely in the final DEMO operation (only one plasma scenario)

→ **but cannot be excluded during DEMO commissioning**

4) Unforeseen events, e.g. impurity events, leakage, SC magnet quench

→ **above a certain size or speed of the event, the control system can only reduce but not avoid machine damage (mitigation)**

→ try to avoid unforeseen events by design (robust components, low wall loads, low amount of accumulated dust)

→ The probability of causes #1 - #3 can be in principle reduced by expensive choices in machine and control design; **for #4 this is less obvious**.

→ DEMO should have a certain degree of resilience against disruptions
<figure><img src="image_43.png" /></figure>

* **QH mode:** results in additional requirements for diagnostics and actuators

* requires control of plasma rotation (NBI as actuator)

* requires measurement of core plasma rotation (high resolution x-ray spectroscopy under at least 20 degree angle against poloidal plane for W or Xe) → 20 km/s should be measurable

* requires also measurement of edge rotation gradient (high resolution imaging x-ray spectroscopy under at least 20 degree angle against poloidal plane for Ar) → 15 km/s should be measurable

<figure><img src="image_44.png" /></figure>

* **I mode:** probably no additional diagnostic and actuator functionality needed but more stringent requirements

* plasma operation under reversed B field → H mode threshold increases

* I mode threshold is lower than H mode threshold → keep $P_{sep}$ between $P_{L1}$ and $P_{LH}$ : $P_{L1} < P_{sep} < P_{LH}$

<figure><img src="image_45.png" /></figure>

* **„Flexi DEMO / advanced scenario“:** probably requires q profile control

* beam aided spectroscopy could be applied in plasma edge region (limited by beam penetration?)

* for core plasma, consider IR polarimetry (additional requirements)

* A lot of NBI or ECRH power needed

<figure><img src="image_46.png" /></figure>

* **Negative triangularity (L mode)**

* Less stringent $P_{sep}$ control needed

* Sightline geometry of some diagnostics to be adapted

<figure><img src="image_47.png" /></figure>