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arxiv: 2604.04751 · v1 · submitted 2026-04-06 · ⚛️ physics.plasm-ph · nucl-ex

Hydrogen Inventory Simulations for PFCs (HISP)

Kaelyn Dunnell , Adria Lleal , Etienne Augustin Hodille , Jonathan Dufour , Remi Delaporte-Mathurin , Tom Wauters This is my paper

Pith reviewed 2026-05-10 19:40 UTC · model grok-4.3

classification ⚛️ physics.plasm-ph nucl-ex
keywords tritium retentionplasma facing componentsITERhydrogen inventorybakingtritium removalFESTIMsimulation
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0 comments X

The pith

Baking removes nearly 88 percent of tritium from ITER's tungsten divertor in HISP simulations, while adding glow discharge conditioning or deuterium pulses changes totals by less than 10 percent.

A machine-rendered reading of the paper's core claim, the machinery that carries it, and where it could break.

The paper introduces HISP, an open-source code that converts plasma exposure data into averaged one-dimensional inputs for hydrogen transport modeling to predict tritium buildup in fusion reactor walls and divertors. It applies the tool to ITER to compare tritium removal methods during and after deuterium-tritium pulses. The results indicate that ten days of operation produce about 35 grams of tritium, mostly trapped in boron layers on the divertor. Baking stands out as the strongest removal technique, cutting inventory sharply in both tungsten and boron, whereas other methods add little once baking is used.

Core claim

HISP converts outputs from plasma codes into spatially averaged inputs for one-dimensional hydrogen transport simulations with FESTIM. When run for ITER's first wall and divertor under three DT operation and removal scenarios, the model yields approximately 35 grams of tritium after 10 days, with nearly 80 percent residing in co-deposited boron layers in the divertor. Baking reduces tritium by almost 88 percent in tungsten and 30 percent in boron there, while glow discharge conditioning peaks at 23 percent removal in the tungsten first wall and low-power deuterium pulses reach 13 percent across the divertor; adding the latter two methods to baking alters final inventories by less than 2% in

What carries the argument

HISP tool that averages plasma code outputs into 1D exposure conditions for FESTIM hydrogen transport models to track isotope inventories across operation and removal scenarios.

If this is right

  • Ten days of DT operation produce roughly 35 g of tritium in first-wall and divertor components, with 80 percent trapped in boron co-deposits.
  • Baking lowers tritium inventory by nearly 88 percent in tungsten and 30 percent in boron within the divertor.
  • Glow discharge conditioning removes up to 23 percent of tritium from the tungsten first wall at peak efficiency.
  • Low-power deuterium pulses remove up to 13 percent of tritium from the entire divertor.
  • Combining glow discharge conditioning and deuterium pulses with baking changes final tritium inventories by less than 2 percent in the first wall and 10 percent in the divertor.

Where Pith is reading between the lines

These are editorial extensions of the paper, not claims the author makes directly.

  • Prioritizing baking in maintenance cycles could keep tritium levels low enough to simplify regulatory compliance for long-pulse fusion devices.
  • If the averaging step holds, the same workflow could be adapted to evaluate removal strategies in other tokamak geometries without full 3D transport codes.
  • Dominance of boron co-deposition implies that surface material selection will strongly influence overall tritium control budgets in future reactors.
  • Extending the simulations to multi-year operation or varied plasma regimes would test whether baking retains its advantage over longer timescales.

Load-bearing premise

Averaging three-dimensional plasma exposure conditions into one-dimensional inputs for the transport model still captures the essential material evolution and co-deposition behavior without major loss of accuracy.

What would settle it

Measurements of actual tritium retained in ITER's divertor tungsten and boron surfaces after a sequence of DT pulses followed by baking that show retention reductions far below the simulated 88 percent for tungsten and 30 percent for boron.

Figures

Figures reproduced from arXiv: 2604.04751 by Adria Lleal, Etienne Augustin Hodille, Jonathan Dufour, Kaelyn Dunnell, Remi Delaporte-Mathurin, Tom Wauters.

Figure 1
Figure 1. Figure 1: Illustration of an ITER Fusion Power Operation Campaign. A 16 [PITH_FULL_IMAGE:figures/full_fig_p001_1.png] view at source ↗
Figure 2
Figure 2. Figure 2: HISP workflow. See Figure [PITH_FULL_IMAGE:figures/full_fig_p003_2.png] view at source ↗
Figure 3
Figure 3. Figure 3: The three two-week scenarios tested in ITER, composed of pulses [PITH_FULL_IMAGE:figures/full_fig_p003_3.png] view at source ↗
Figure 4
Figure 4. Figure 4: ITER PFC surfaces represented by HISP bins, here plotted as seg [PITH_FULL_IMAGE:figures/full_fig_p004_4.png] view at source ↗
Figure 5
Figure 5. Figure 5: Illustration of heat load maps on FW panels (from bottom to top: row 3, 4, and 5) used for assigning wetted (colored) and shadowed (gray) surface areas and averaged loads to the 1D bin simulations of HISP. The maps result from SMITER analysis by F. Fernández-Marina of a plasma baseline scenario with ∆rSEP = 10 cm, using the panel shaping of the Be wall. Wetted areas re￾ceive both atom and ion fluxes, while… view at source ↗
Figure 7
Figure 7. Figure 7: Temporal evolution of the tritium inventories and temperature for [PITH_FULL_IMAGE:figures/full_fig_p007_7.png] view at source ↗
Figure 6
Figure 6. Figure 6: Evolution of tritium inventory in the FW and divertor at the end of scenarios A, B, and C. The trapped and mobile tritium inventories in tungsten are described in [PITH_FULL_IMAGE:figures/full_fig_p007_6.png] view at source ↗
Figure 8
Figure 8. Figure 8: Temporal evolution of the tritium inventories and temperature for [PITH_FULL_IMAGE:figures/full_fig_p008_8.png] view at source ↗
Figure 10
Figure 10. Figure 10: Temporal evolution of the hydrogen inventories and temperature for [PITH_FULL_IMAGE:figures/full_fig_p008_10.png] view at source ↗
Figure 9
Figure 9. Figure 9: Temporal evolution of the hydrogen inventories and temperature for [PITH_FULL_IMAGE:figures/full_fig_p008_9.png] view at source ↗
read the original abstract

Hydrogen Inventory Simulations for Plasma facing components (HISP) is an open-source simulation tool to model the evolution of hydrogen (H) isotopes inventory in plasma-facing-components (PFCs) of magnetic confinement fusion devices. The objective was to produce a demonstrative study describing the efficiency of tritium (T) removal strategies in ITER. HISP transforms plasma code outputs to spatial-averaged inputs along ITER's first wall (FW) and divertor for 1D H transport models using FESTIM. Exposure conditions were tested in three scenarios that included DT operation and varied T removal methods. Generally, DT operation resulted in $\approx$ \SI{35}{g} of T in FW and divertor components after 10 days of DT pulses. Almost \SI{80}{\%} of the total T inventory resided in co-deposited boron layers in the divertor. Baking proved to be the most effective T removal method in the divertor, decreasing T inventory by almost \SI{88}{\%} for tungsten and almost \SI{30}{\%} for boron. T removal was also evaluated from Glow Discharge Conditioning (GDC) - with a peak efficiency of \SI{23}{\%} in the tungsten FW - and low power deuterium (DD) pulses - with a peak efficiency of \SI{13}{\%} in the entire divertor. Due to the high removal efficiency of baking, inclusion of GDC and DD pulses in the tested scenarios did not meaningfully change final T inventory values, which varied by less than \SI{2}{\%} in the FW and \SI{10}{\%} in the divertor between scenarios.

Editorial analysis

A structured set of objections, weighed in public.

Desk editor's note, referee report, simulated authors' rebuttal, and a circularity audit. Tearing a paper down is the easy half of reading it; the pith above is the substance, this is the friction.

Referee Report

3 major / 2 minor

Summary. The paper introduces the open-source HISP tool, which converts outputs from plasma codes into spatially averaged 1D inputs for FESTIM hydrogen transport simulations to model tritium (T) inventory evolution in ITER plasma-facing components. After 10 days of DT operation, it reports ~35 g total T inventory (~80% in boron co-deposits in the divertor) and evaluates removal strategies, finding baking most effective (88% reduction in tungsten, 30% in boron) while GDC and DD pulses add little benefit due to baking's dominance, with final inventories varying <2% in the first wall and <10% in the divertor across scenarios.

Significance. If the modeling assumptions hold, the work provides a practical, open-source framework for forward simulation of T retention and removal in ITER-relevant conditions using established FESTIM code, yielding concrete quantitative estimates that could inform PFC operational planning and T inventory management in magnetic confinement fusion. The emphasis on reproducible tool development and scenario testing is a strength.

major comments (3)
  1. [model setup / input preparation] The spatial averaging of 3D plasma code outputs (fluxes, temperatures, particle energies) into representative 1D inputs for FESTIM along the first wall and divertor (described in the model setup and input preparation) does not address how this procedure handles nonlinear hydrogen transport, trapping, and co-deposition processes. Since retention rates depend on local concentration and temperature, averaging high-flux strike-point or leading-edge zones risks underestimating peak inventories and overstating removal efficiencies such as the reported 88% baking reduction in tungsten.
  2. [results / quantitative outcomes] The headline quantitative results (~35 g total T, 80% in boron, 88% and 30% baking reductions, <2-10% variation from added GDC/DD) are presented without sensitivity studies, error bars, or propagation of uncertainties from plasma code inputs and FESTIM material parameters (diffusion, trapping, recombination coefficients). This leaves the specific claims difficult to assess for robustness.
  3. [results / discussion] No validation of the overall HISP/FESTIM model or the specific parameter choices against experimental T inventory data from tokamaks or ITER-relevant benchmarks is provided, despite the concrete numerical outcomes reported in the abstract and results.
minor comments (2)
  1. [methods] Clarify the exact spatial averaging procedure (e.g., how area-weighted means are computed and over what poloidal segments) to allow reproducibility.
  2. [abstract / results] The abstract and results would benefit from explicit statements on the time scales of the DT pulses and baking steps to contextualize the 10-day operation period.

Simulated Author's Rebuttal

3 responses · 0 unresolved

We thank the referee for their constructive and detailed comments on the HISP manuscript. We have addressed each major point below, making revisions where the concerns are valid and can be incorporated without altering the core scope of this demonstrative study.

read point-by-point responses

  1. Referee: The spatial averaging of 3D plasma code outputs (fluxes, temperatures, particle energies) into representative 1D inputs for FESTIM along the first wall and divertor (described in the model setup and input preparation) does not address how this procedure handles nonlinear hydrogen transport, trapping, and co-deposition processes. Since retention rates depend on local concentration and temperature, averaging high-flux strike-point or leading-edge zones risks underestimating peak inventories and overstating removal efficiencies such as the reported 88% baking reduction in tungsten.

    Authors: We agree that spatial averaging into 1D profiles can smooth over local nonlinear effects in high-flux zones, which may affect peak local inventories and potentially overstate average removal efficiencies. The HISP tool uses this standard approach to enable whole-device 1D simulations at manageable computational cost. In the revised manuscript we will add a new paragraph in the Discussion section explicitly addressing this limitation, including a qualitative estimate of how strike-point peaks might influence the reported global inventory and baking reduction figures. revision: partial

  2. Referee: The headline quantitative results (~35 g total T, 80% in boron, 88% and 30% baking reductions, <2-10% variation from added GDC/DD) are presented without sensitivity studies, error bars, or propagation of uncertainties from plasma code inputs and FESTIM material parameters (diffusion, trapping, recombination coefficients). This leaves the specific claims difficult to assess for robustness.

    Authors: We acknowledge that the quantitative claims would be more robust with explicit sensitivity analysis and uncertainty estimates. In the revised version we will add a dedicated subsection performing sensitivity studies on key FESTIM parameters (diffusion, trapping energies, recombination coefficients) and on variations in the plasma-code input fluxes. This will include reported ranges for the ~35 g inventory and the removal efficiencies. revision: yes

  3. Referee: No validation of the overall HISP/FESTIM model or the specific parameter choices against experimental T inventory data from tokamaks or ITER-relevant benchmarks is provided, despite the concrete numerical outcomes reported in the abstract and results.

    Authors: FESTIM has been validated against experimental retention data in several prior publications (we will add the relevant citations to the Methods section). The material parameters used here are taken from the established literature. Because ITER is not yet operating, direct experimental benchmarks for the full ITER scenario do not exist; the present work is therefore a forward demonstration rather than a validated prediction. We will expand the Introduction and Discussion to make this distinction and the reliance on validated sub-models clearer. revision: partial

Circularity Check

0 steps flagged

No circularity: forward simulation from external plasma codes and FESTIM

full rationale

The paper computes T inventories and removal efficiencies as direct outputs of 1D FESTIM transport simulations driven by spatially averaged inputs from separate plasma codes. No parameters are fitted to the reported inventories, no self-citation chain justifies the central claims, and the derivation does not reduce any result to a definition or renaming of its own inputs. The spatial-averaging step is an explicit modeling choice whose validity can be assessed independently of the numerical outputs.

Axiom & Free-Parameter Ledger

1 free parameters · 1 axioms · 0 invented entities

The central results rest on standard assumptions of 1D transport modeling and spatial averaging of plasma conditions; no new physical entities are postulated and the only free parameters are those already present in the FESTIM code.

free parameters (1)
  • FESTIM material parameters (diffusion, trapping, recombination coefficients)
    1D hydrogen transport models require these coefficients, which are typically taken from literature or fitted to separate experiments.
axioms (1)
  • domain assumption Spatial averaging of plasma code outputs yields representative boundary conditions for 1D FESTIM runs along the first wall and divertor
    The abstract states that plasma outputs are transformed to spatial-averaged inputs for the 1D models.

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