A framework for the assessment and control of ITER main chamber heat loads

Anand, H.; Pitts, R.A.; Vries, P. C. de; Snipes, J. A.; Kos, L.; Gribov, Y.; Zabeo, L.; Nunes, I.; Brank, M.

doi:10.1088/1741-4326/ab6150

Cited by 11 publications

(9 citation statements)

References 30 publications

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“…Figure 10(a) depicts the stationary q ⊥ distribution for a typical 15 MA, ELM-y H-mode flattop for ITER on undamaged FWPs #7-9 analyzed in [22,23]. Heat loads include a stationary (inter-ELM) component and an averaged component for mitigated ELMs.…”

Section: Smiter Analysis Of Damaged Fwpsmentioning

confidence: 99%

Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events

Coburn¹,

Lehnen²,

Pitts³

et al. 2021

Nucl. Fusion

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An analysis workflow has been developed to assess energy deposition and material damage for ITER vertical displacement events (VDE) and major disruptions (MD). This paper describes the use of this workflow to assess the melt damage to be expected during unmitigated current quench (CQ) phases of VDEs and MDs at different points in the ITER Research Plan. The plasma scenarios are modelled using the DINA code with variations in plasma current Ip, disruption direction (upwards or downwards), Be impurity density nBe, and diffusion coefficient χ. Magnetic field line tracing using SMITER calculates time-dependent, 3D maps of surface power density q_⊥ on the Be-armored first wall panels (FWP) throughout the CQ. MEMOS-U determines the temperature response, macroscopic melt motion, and final surface topology of each FWP. Effects of Be vapor shielding are included. Scenarios at the baseline combination of Ip and toroidal field (15 MA/5.3 T) show the most extreme melt damage, with the assumed nBe having a strong impact on the disruption duration, peak q_⊥ and total energy deposition to the first wall. The worst-cases are upward 15 MA VDEs and MDs at lower values of nBe, with q_(⊥,max)=307 MW/m^2 and maximum erosion losses of ~2mm after timespans of ~400-500 ms. All scenarios at 5 MA avoided melt damage, and only one 7.5 MA scenario yields a notable erosion depth of 0.25 mm. These results imply that disruptions during 5 MA, and some 7.5 MA, operating scenarios will be acceptable during the Pre-Fusion Power Operation phases of ITER. Preliminary analysis shows that localized melt damage for the worst-case disruption should have a limited impact on subsequent stationary power handling capability.

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Section: Smiter Analysis Of Damaged Fwpsmentioning

confidence: 99%

Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events

Coburn¹,

Lehnen²,

Pitts³

et al. 2021

Nucl. Fusion

Self Cite

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“…For each equilibrium, weighting factors associated with the magnitudes of the peak heat flux are extracted using SMITER for implementation into the RT 2D model-based approach. Further details about the general methodology can be found in [16].…”

Section: Heat Flux Density and Surface Temperature Descriptorsmentioning

confidence: 99%

“…The framework presented in [16] is also extended to estimate the surface temperature on the DIII-D IWL. The surface temperature can be determined from q ⊥,PFC with the use of a thermal model comprising of a 1D thermal diffusion model.…”

Section: Heat Flux Density and Surface Temperature Descriptorsmentioning

confidence: 99%

“…Probe λ q,near (mm) λ q,main (mm) R q q 0,main (MW m with q 0,main = B φ,OMP P SOL 4πR OMP (λ q,main + R q λ q,near )B θ,OMP , R q ≡ q 0,near q 0,main (16) where B φ,OMP and B θ,OMP are the toroidal and poloidal component of the magnetic field at the OMP. An example of the resulting q dep (φ, Z) 2D map from IR thermography, along with the limiter plasma equilibrium used in the study is shown in figures 1(a) and (b).…”

Section: Shotmentioning

confidence: 99%

“…Recently, an RT 2D model-based control oriented heat flux assessment system accounting for the effect of 3D PFCs was successfully developed for the ITER plasma control system (PCS) [16]. This paper describes the adaptation and experimental demonstration of the RT heat flux algorithm developed for ITER, on an operating device-the DIII-D tokamak [17].…”

Section: Introductionmentioning

confidence: 99%

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Model-based real-time surface heat flux and temperature estimation for the DIII-D tokamak

Anand¹,

Eldon²,

Humphreys³

et al. 2021

Nucl. Fusion

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A control-oriented model for monitoring of wall power flux densities on the DIII-D tokamak has been successfully implemented and validated experimentally. Future reactors will have to withstand severe steady state high heat flux loads on plasma-facing components (PFCs). Due to the difficulty of directly-measuring local heat fluxes on these components, monitoring and protection of PFCs during the plasma discharge can benefit from simplified physics-based real-time functional models to estimate and guide heat load control. As a first step into the development, a control-oriented model for monitoring of wall power flux densities and temperatures on DIII-D tokamak has been successfully implemented. The paper discusses the experimental demonstration and comparison of the 2D model-based wall heat flux algorithm on the DIII-D inner wall limiter (IWL) against infra-red (IR) camera heat flux measurements for limited plasma configurations. The paper also reports on the benchmarking of the field line tracing environment, SMITER, developed at ITER organization on DIII-D tokamak against experimental IR diagnostic data and the derivation of the component shaping weighting factors for the 2D model-based approach. Extension of the model-based approach for surface temperature estimation on the DIII-D IWL is also presented.

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Assessment of the Impact of Magnetic Equilibria Designs on the Stationary Plasma Heat Flux Deposition onto the CFETR First Wall

Nian

Yang

et al. 2023

J Fusion Energ

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A framework for the assessment and control of ITER main chamber heat loads

Cited by 11 publications

References 30 publications

Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events

Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events

Model-based real-time surface heat flux and temperature estimation for the DIII-D tokamak

Assessment of the Impact of Magnetic Equilibria Designs on the Stationary Plasma Heat Flux Deposition onto the CFETR First Wall

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