2nd Fusion Neutronics Meeting 2026

8 Jun 2026, 09:00 → 12 Jun 2026, 19:30 Europe/Berlin

FTU (Karlsruhe Institute of Technology, Campus north)

Dieter Leichtle

We are delighted to invite you to the 2^nd Fusion Neutronics Meeting, which will take place from June 8–12, 2026, during the pleasant early summer season at the KIT Campus North in Karlsruhe. 

 

This workshop continues the long-standing tradition of the ITER Neutronics Meetings series, which was hosted at KIT in 2016, exactly ten years ago. Now we expand the scope to encompass a broader perspective on fusion neutronics, inspired by the rapid advancements in diverse fusion concepts and the emergence of private fusion industry initiatives. These include tokamaks, stellarators, laser-ignited fusion, and other innovative fusion approaches, as well as dedicated fusion neutron source facilities and blanket validation/test facilities. The workshop will cover a wide range of topics in fusion neutronics, including but not limited to simulation workflows and modelling, neutron transport, activation analysis, shielding design, and shutdown dose rates. In addition, it will address key aspects such as simulation tools, nuclear data libraries, and nuclear experiments specifically dedicated to benchmarking and validation. 

 

This workshop aims to serve as a valuable opportunity to foster and expand the fusion neutronics community. It will bridge the growing private fusion sector with established public research efforts, facilitating the exchange of expertise and know-how, and establishing and reinforcing connections within the global fusion neutronics community. Furthermore, it will strengthen collaborations on open-source simulation codes, openly available nuclear data libraries, and evaluation efforts.  

 

We would love to welcome you with opening talks on the history of fusion neutronics, interesting scientific talks and social events, technical tours of key KIT scientific facilities, lectures on codes and tools, and enjoyable sunny weather in Karlsruhe with local German beers! See you in Karlsruhe! 

Abstract_template.docx

Monday 8 June

09:00 → 10:30 OpenMC Workshop: OpenMC code presentation and learning course

The workshop on the OpenMC radiation transport code for the Fusion Neutronics applications is organized by the OpenMC developers. It will be presented by Dr. Jonathan Shimwell from the Proxima Fusion company.

10:30 → 11:00 Coffee Break

11:00 → 12:00 OpenMC Workshop: OpenMC Users' Feedback and Discussion

The workshop on the OpenMC radiation transport code for the Fusion Neutronics applications is organized by the OpenMC developers. It will be presented by Dr. Jonathan Shimwell from the Proxima Fusion company.

12:00 → 13:00 Lunch for the OpenMC Workshop participants

13:00 → 13:30 Welcome

13:30 → 14:30 Invited Session

13:30 Fusion Neutronics: Highlights from Fifty Years of Development

The development of Fusion Neutronics is closely related to Fusion Technology (FT) kicked-off in the 1970ies with the first tokamak reactor studies conducted then at the University of Wisconsin, Madison, in the US. Such studies necessitate the application of suitable computational tools, models and nuclear data to provide the nuclear responses required for assessing the nuclear performance of the considered reactor, in particular the tritium breeding, the nuclear power production and the radiation shielding capabilities. The underlying neutronic calculations were quite simple in those early times of FT, mainly based on one-dimensional S(N)-calculations using nuclear data from the ENDF/B-IV data file which already included some cross-section data for 14-MeV neutrons.
In the following decades a large number of power reactor studies were then conducted, notably in the US and in the EU with the European Fusion Technology Programme launched in 1983. Starting with the 1980s studies, the simple one-dimensional modelling approach was gradually replaced by the Monte Carlo method for the neutronics calculations. This enabled a more realistic modelling of the investigated reactor configurations and thus provided more accurate results for the nuclear responses, in particular the tritium breeding capability.
While the code development was mostly performed outside the fusion community, fusion specific activities were conducted in the experimental field with a series of integral 14-MeV neutron experiments for validating the neutronics calculations, but also differential neutron cross-section measurements in the fusion relevant energy around 14 MeV and beyond. In addition, dedicated efforts were conducted to establish qualified nuclear data libraries such as the Fusion Evaluated Nuclear Data Library (FENDL) of the IAEA.
With the launch of the ITER project, in the late 1989ies, and the IFMIF project for a high intense neutron source in the mid 1990ies, both with the objectives to build and operate the facilities, it became necessary to perform well qualified neutronics calculations satisfying also the safety requirements. This necessitated, among other, a very accurate geometry modelling based on detailed engineering CAD models and suitable methods for shut-down dose-rate assessments. ITER thus acted as driving force for related dedicated code and method developments.

The presentation briefly reviews the major developments in the field of fusion neutronics and highlights specific achievements in the computational and experimental area.

Speaker: Dr Ulrich Fischer (Retired, ex-KIT)

Abstract_UFischer_FNW2026_Invited (final).docx

14:00 Status of ITER Neutronics

G. Náfrádi1, S. Puthanveetil1, E. Polunovskiy1, G. Zeng1, G. Pedroche2, P. Martinez2, P. Guijosa2, N. Khvatkin2, A. Mayo2, L. Ruiz2, J. Alguacil2, V. López2, P. Sauvan2, R. Juárez2, R. Pampin3, D. Laghi3, M. Campos Fornés3, A. Bittesnich3, A. Cubí3, M. Di Giacomo3, M. Fabbri3, A. Kolšek3, T. Schioler1

1 ITER Organization, Route de Vinon-sur-Verdon - CS 90 046 - 13067 St Paul Lez Durance Cedex – France
2 Universidad Nacional de Educación a Distancia (UNED), C/Juan del Rosal 12, Madrid, Spain
3Fusion for Energy (F4E), Josep Pla 2, Barcelona 08019, Spain
Corresponding Author Email: subhash.puthanveetil@iter.org

This presentation aims to provide a comprehensive status update on ITER global radiation maps and ongoing neutronics related activities. These maps will serve multiple critical functions, including safety assessments, design optimization, and qualification processes. At present, however, the focus is on supporting the upcoming RPrS update.
The presentation will outline the first results of mode-0 best estimate maps, with special focus on uncertainty analysis. It will also give an overview of ongoing and future simulation tasks, including decay heat assessments, dose rates at the site boundary, dust and radwaste estimates, and maps for SRO, mode-1 and mode-2.

Speaker: G. Náfrádi

Status_of_ITER_Neutronics_G3QHZR_v1_0 (2).docx

14:30 → 15:30 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

14:30 The latest release of ITER Tokamak model: E-lite_R250630

In 2020, a high-fidelity 360° MCNP model of the ITER tokamak was developed by assembling seven replicas of the C-Model (a 40° sector of a regular vacuum vessel segment), integrating the 80° toroidal segment of the Neutral Beam Injector (NBI), and incorporating all available MCNP models of ITER ports, followed by targeted refinements. The resulting E-lite model represented a significant advancement in nuclear analysis, enabling unprecedented realism in the assessment of critical parameters, and a paper on the work was subsequently published in Nature Energy.

Following the initial release, the model underwent targeted updates for specific applications, though no formal versions were issued. At the 34th Meeting of the ITER Council in 2024, the ITER Organization proposed an updated project baseline to the Council. The revised baseline focuses on accelerating the start of major research operations by optimizing tokamak assembly, strengthening pre-assembly testing, and reducing assembly and commissioning risks. As part of licensing considerations, the beryllium used as the plasma facing material of the blanket First Walls (FW) was also replaced by tungsten, a change expected to impact radiation conditions both inside the plasma chamber and outside the Vacuum Vessel (VV). This milestone presented an ideal opportunity to consolidate all updates from recent years, implement the FW material change, and produce a new official release of the E-lite model – E-lite_R250630.

In this work the latest update of the E-lite model is presented, featuring a heterogeneous (tungsten FW, explicit water channels and material layers separated) representation of most Blanket Shield Modules (BSM) and introducing approximately 660 tons of components within the Bioshield. The presentation also highlights additional improvements to the ITER machine representation, alongside the improved modularity of E-lite to simplify and automatise the extraction of specific sectors.

This model was tested using the Gitronics workflow, a novel modular and Git-based approach for the management and development of complex radiation transport models. In a pilot study, E-lite_R250630 was decomposed into smaller units (such as the envelope structure, universes, and materials), stored in a Git repository, and reassembled using tools within Gitronics. The results demonstrated statistical equivalence with the original E-lite model, unlocking potential for future neutronics reference model development, and shown a fast and reliable deployment of multiple, operation-specific E-lite configurations (i.e., SRO, DT1, DT2).

Speaker: Aljaz Kolsek (Fusion for Energy (F4E))

E-lite_R250630_v1.1 - Aljaz_Kolsek.pdf

14:50 Best Estimate ITER Mode-0 Radiation Maps

ITER will be a major step towards demonstrating the scientific and technological feasibility of nuclear fusion as a clean and safe energy source. It will operate the largest Tokamak ever built. The radiation fields expected in ITER, both during and after operation, make nuclear analyses particularly valuable. Some of these analyses are expected to support the update of ITER Rapport Préliminaire de Sûreté (RPrS), the safety report periodically updated and reviewed by the French Nuclear Safety Authority (ASNR). The generation of 3D radiation maps is expected to significantly support the demonstration of ITER’s compliance with the French radiological safety regulations, aimed at protecting workers and the public.
In this work, we present the Best Estimate (BE) ITER Mode-0 Radiation Maps that provide the radiation levels (neutron flux, photon flux, biological dose rate, absorbed dose in silicon, etc) inside the Tokamak Complex during machine operation. Both the plasma and the Tokamak Cooling Water System (TCWS) activated water were the radiation sources considered.
These radiation maps represent a significant upgrade in terms of robustness with respect to previous releases. They are the first ones calculated in a single MCNP model of the whole ITER facility, the BE ITER full model. This model represents the tokamak and the site buildings, avoiding the previously existing need of coupling separate models through intermediate radiation sources that introduced unassessed uncertainties in the results. In addition, the TCWS activated water source was updated in accordance with its latest design.

Speaker: Gabriel Pedroche (UNED)

02_Abstract_Mode0Maps_GPE_PVS.docx

15:10 Best Estimate ITER Full Model for radiation safety demonstration

ITER will be a major step towards demonstrating the scientific and technological feasibility of nuclear fusion as a clean and safe energy source. It will operate the largest Tokamak ever built. The radiation fields expected in ITER, both during and after operation, make nuclear analyses particularly valuable. Some of these analyses constitute a significant part of the ITER Rapport Préliminaire de Sûreté (RPrS), the safety report periodically updated and reviewed by the French Nuclear Safety Authority (ASNR). The generation of 3D radiation maps is required to demonstrate ITER’s compliance with the French radiological safety regulations, aimed at protecting workers and the public.

With this perspective, since 2024 the UNED team has been working on the production of a new MCNP representation of the ITER facility to support the upcoming radiation maps within the next RPrS update. The produced model addresses a robustness requirement. A remarkable aspect of it is that it leaves behind the common practice of coupling models. This coupling resulted in the introduction of unassessed uncertainties due to source binning, weakening the robustness of the results.

Motivated by this need, and in close collaboration with the ITER Organization, the UNED team produced the first integral Best Estimate (BE) MCNP representation of the ITER facility: the BE ITER Full Model. This model represents the tokamak, the site buildings and the tokamak cooling water systems. The model was produced without deliberate pessimism and gradually reducing legacy conservatism embedded in earlier ITER reference models. It incorporates a block structure architecture, a desired feature in many previous ITER reference models. It enables the modularity of the model: the extraction of partial representations using dedicated tools, thereby supporting future local studies avoiding ad hoc approaches based on producing local models. This architecture also facilitates version control and promotes the use of the model as a geometry database by the broader community. Moreover, it enables the deployment of uncertainty quantification schemes.

The BE ITER Full Model will increase the credibility and robustness of nuclear analysis results. It is the cornerstone of the ITER radiation safety case and a significant improvement regarding fusion safety demonstrations. Finally, it was produced with the past, present, and future in mind. It lays the groundwork for next steps, including its own Gitronization and progress toward a digital twin of ITER’s operating fusion reactor.

Speaker: Pol Guijosa (UNED)

Abstract_FNM26_PGA_final.docx

15:30 → 16:00 Coffee Break

16:00 → 17:00 Overviews and Neutronics Strategies

16:00 Overview of the UKAEA Neutronics Research Programme

The Applied Radiation Technology group at the United Kingdom Atomic Energy Authority (UKAEA) comprises a group of experts in neutronics and radiometrics to support the delivery of fusion energy. The group has expanded significantly in recent years, driven by unprecedented investment in the UK fusion programme and the rapid growth of the international private fusion sector. This growth is underpinned by decades of cumulative expertise in the development and application of advanced methods for modelling and measurement of radiation fields.

This presentation provides an overview of the group’s broad research portfolio, with particular focus on recent developments and applications in neutronics. A major challenge in radiation transport analysis lies in the defeaturing, simplification, and conversion of CAD models. To address this, UKAEA has developed MCIO, a tool for conversion of CAD geometry to constructive solid geometry (CSG), currently supporting MCNP, OpenMC and Serpent. As well as the underlying conversion algorithms, the development has focused on optimized void generation and conversion performance gains for large-scale models achieved through multi-processing. Through collaboration with EUROfusion partners, UKAEA continues to advance the modelling of fluid activation. The GammaFlow code is being extended to incorporate aspects of CFD, enabling more accurate prediction of fluid flow and thereby the gamma source terms from activated fluids. A further research area addresses the modelling of Activated Corrosion Products (ACPs). Recognising both the safety significance of this phenomenon and the lack of suitable tools worldwide, UKAEA is developing TRACTOR, a new code designed to model pipe corrosion and the transport of activated particulates.

Ensuring the credibility of nuclear analysis remains a core priority, with substantial effort devoted to validation and verification (V&V) of radiation transport codes and nuclear data. In collaboration with F4E, UKAEA is developing the open-source tool JADE, with a strong focus on OpenMC integration. The automated V&V datasets enabled by JADE will contribute to unifying ongoing international benchmarking activities and the broader adoption of this code. As fusion approaches deployment, robust neutronics‑based technical evidence is needed to support safety, security, and non‑proliferation policy, building on ongoing and recent UKAEA engagement with international partners, including through secondments to the IAEA. The presentation concludes with examples of nuclear analysis supporting private fusion, demonstrating applications to stellarators, inertial confinement, and magneto-inertial fusion concepts, often requiring advanced or novel extensions to core tools and methods.

Speaker: Alex Valentine (UKAEA)

Abstract_FNM_Valentine_v1.0.pdf

16:20 Empowering Neutronics Fusion Development and Excellence: F4E’s Approach to Knowledge, Open-Source Tools, and Industry Collaboration

As per title.

Speaker: Marco Fabbri (Fusion For Energy)

Empowering Neutronics Fusion Development and Excellence at F4E.docx

16:40 Overview of the neutronics studies for the Divertor Tokamak Test facility

The Divertor Tokamak Test (DTT) facility will be constructed at the ENEA Frascati Research Centre to address the problem of power exhaust, one of the key challenges of the future fusion power plants. DTT is a medium-sized superconducting tokamak operating with a deuterium plasma and with a substantial level of external heating power, to reproduce the divertor heat loads foreseen for ITER and DEMO. During the high-performance phases, DTT will produce 3∙10$^{22}$ neutrons per year (1.5∙10$^{17}$n/s) from D-D fusion reactions. In addition, the generation of a non-negligible amount of 14 MeV neutrons from D-T reactions is expected due to the triton burn-up. As a consequence, the machine components will be exposed to an intense neutron and gamma irradiation, as well as a high neutron-induced activation. This has a significant impact on the design of the tokamak components, auxiliary heating and diagnostics systems, as well as on licensing, maintenance, decommissioning and waste management.
In this work an overview of the neutronics studies supporting the DTT design will be reported, with a focus on two particular aspects, the impact of material impurities and the development of a detailed neutronics model. Some components are currently being put out to tender, and challenges may arise during manufacturing. These include inconsistencies in requirements flow-down, the need to rely on commercial off-the-shelf and standard components made from uncertified materials, the lack of established testing methods and protocols, and the increased costs and management complexity associated with certified purpose-built materials. To better address these issues, sensitivity studies have been carried out to assess how the increase in material impurity levels affects shutdown dose rate and radiological impact on working environment and personnel. A novel DTT neutronics model is under development by the neutronics team to enable a high-fidelity description of the components following the finalization of the DTT design. The development of the DTT neutronics model will be presented, also in the light of a parallel implementation of MCNP and OpenMC models.

Speaker: Flammini Davide (ENEA)

FNM_abstract_Flammini_final.docx

17:00 → 18:00 Poster Session

17:00 14-MeV Neutron Irradiation Experiments for Fusion Applications

A comprehensive series of neutron irradiation experiments were carried out at the Institute for Plasma Research using a 14-MeV neutron generator to support fusion technology development, nuclear data validation, and material qualification under high-radiation environments. Key investigations included neutron-induced reaction cross-section measurements for isotopes of Sb, Zr, Sn, Sr, and Re at a neutron energy of ~14.96 MeV using activation techniques. Reactions such as ⁹⁰Zr(n,p), ¹²¹Sb(n,2n), ¹¹⁷Sn(n,p), 118Sn(n,α), 120Sn(n,α), ⁸⁶Sr(n,2n), and ¹⁸⁵Re(n,2n) were studied with detailed uncertainty quantification using covariance analysis, and the measured data were compared with predictions from nuclear model codes including TALYS-2.0, EMPIRE-3.2.3, and TASMAN-2.0 to address discrepancies in legacy datasets. In parallel, neutron irradiation experiments were conducted to qualify ITER-relevant materials and components under 14 MeV neutron exposure, including a stepper motor system for ITER CXRS pedestal diagnostics, which demonstrated stable operation, precise positioning, and repeatable performance up to a neutron fluence of ~3.5×10¹² n/cm² under load conditions. Additionally, catalyst materials developed for the ITER hydrogen mitigation system and structural steel alloys were irradiated to evaluate neutron-induced effects and attenuation characteristics, while a conceptual neutron moderator design was developed using MCNP simulations to tailor the neutron energy spectrum for sensor qualification. Overall, the results show good agreement with theoretical predictions for most reactions while resolving discrepancies in existing data, and confirm the robustness and functional reliability of ITER-relevant components under neutron irradiation, supporting their suitability for fusion reactor applications.

Speaker: Dr Sudhirsinh Vala (Institute for Plasma Research)

Abstract_Fusion Neutronics Meeting_1.pdf

17:00 Addition of a 1D multi-group neutron diffusion model to the fusion systems code PROCESS

Reactor system codes use reduced-order models to represent each key subsystem of the fusion power plant, such that rapid self-consistent evaluation of full fusion power plant concepts can be performed. Traditionally system codes must rely on fitted empirical data to estimate neutron transport and deposition. PROCESS is the primary systems code used by UKAEA, which could benefit from the addition of a semi-analytical neutronics model. Finding the solution to the full Boltzmann transport equation for neutrons is expensive, therefore a diffusion approximation is applied, and the geometry is simplified to only account for variation in materials and neutron flux in the radial direction. Materials are assumed to be layered perpendicular to the direction at which neutrons diffuse out of the plasma, and the multi-group neutron flux is solved for all points along the radial direction, such that neutron heating, tritium breeding rate, neutron leakage rate, and neutron damage can be calculated.

This model is applied onto a generic fusion reactor’s geometry, and then evaluated against an OpenMC simulation of an analogous geometry. The discrepancies between the two are explored, and the effect on reactor design optimization is discussed. Ultimately, the addition of this neutronics model allows for more realistic constraints and optimization objectives to be calculated: for example, PROCESS can then be ran with a constraint on the coolant pumping power such that the heat removal rate is at least equal to the neutron heating rate; or optimized for a lower DPA on the central solenoid such that the reactor’s down time may be minimized.

Speaker: Mr Christopher Ashe (UKFE)

ocean-wong-process-neutronics 3.docx

17:00 AI-based Program for Nuclear Design and Safety Evaluation - TopMC

To provide intelligent design and analysis in the fields of radiation shielding, material activation, and dose calculation, TopMC (multi-functional program for neutronics calculation, nuclear design and safety evaluation) has been developed. As an updated and extended version of SuperMC, TopMC focus on full-function and high-efficiency neutronics calculation, CAD/image-based accurate modeling for complex irregular geometry, data analysis based on multi-dimensional / multi-style visualization, and especially support intelligent dose calculation and safety evaluation based on Artificial Intelligence (AI). The integrated AI prediction, multi-objective optimization and particle transport model modification modules are included.

The AI prediction module provides various surrogate models based on artificial neural network and mathematical statistics. After training by suitable data, they can replace or couple with the traditional Monte Carlo particle transport programs to perform flux or dose calculations (predictions) in the process of optimization, which accelerate the overall design and optimization work. The multi-objective optimization module automatically sample the different parameters according to the design constraints of the Monte Carlo transport model set up by users through the graphical user interface (GUI). After analyzing the results from AI predictions or transport simulations, this module will provide an optimization scheme for the next batch of models to be calculated, which include geometry dimensions, materials, etc. The particle transport model modification module will locate and modify the parameters in Monte Carlo transport model automatically according to the optimization results without human interference. After verification, a complete new model will be constructed for the next optimization. This modification module can be applied for transport models of different fields according to the parameters ranges specified by users through the GUI. The data managing module will ensure users to perform the whole design and optimization process, and managing their special model and data, which liberate users from coding, manual analysis of simulation results and manual modification of transport models.

The nuclear design and safety evaluation based on AI has been applied in the neutronics analysis for an HCPB DEMO has been accomplished, in which the nuclear performance including neutron wall loading, tritium breeding ratio, nuclear heating and radiation loads was calculated. Results demonstrated its enhanced efficiency in nuclear design and safety evaluation.

Speaker: Shanqi Chen (1)SuperSafety Science & Technology Co., Ltd., Hefei 230088, China; 2)International Academy of Neutron Science, Chongqing, 401331, China)

Abstract_FNM2026-TopMC.pdf

17:00 Characterization of the irradiation cavity of the GENeuSIS Neutron Assembly

Current neutron irradiation facilities cannot fully reproduce the complex radiation spectra expected in fusion tokamaks such as ITER. As a result, experimental qualification of components under fully representative ITER conditions is not yet achievable with conventional test infrastructures.

To address this gap, GENeuSIS (General Experimental Neutron System Irradiation Station) was developed as a modular, flexible, and transportable setup composed of layers of different materials. Its goal is to tailor neutron energy spectra within a dedicated test cavity to replicate radiation environments relevant to specific ITER locations. The system currently relies on 14 MeV neutrons from the Frascati Neutron Generator (FNG).

Two configurations were defined based on prior MCNP simulations: GENeuSIS-I, representing the ITER Port Interspace, and GENeuSIS-II, designed for the Port Cell of Equatorial Port #12. GENeuSIS-II is now fully assembled and operational at FNG. This presentation focuses on experimental characterization of the irradiation cavity using both active and passive neutron detectors, aiming to measure neutron flux and validate the achieved spectral conditions.

Speaker: Marta Damiano (tor vergata university)

abstarct neutronics meeting 2026.docx

17:00 Closed-Form and Machine-Learned Surrogates for Determining Impurity Limits and Operating Windows in Low-Activation Fusion Materials

Deuterium-tritium fusion is often described as intrinsically low-waste, however ensuring that fusion components meet low-level waste (LLW) acceptance criteria under U.S. regulatory frameworks (10 CFR 61) requires coordinated control of impurity content and operating conditions. This study develops an accelerated surrogate framework to support that task by linking isotope-resolved waste disposal rating (WDRᵢ) to impurity concentration, major radius, fusion power, and operating time across a generic tokamak parameter sweep. A closed-form ordinary differential equation (ODE) for WDRᵢ is fit to activation-derived reference data at 100 years cooling, with the goal of enabling rapid estimation of the operating characteristics required to keep the total waste disposal rating from contributing isotopes (WDR) below unity, or conversely, the impurity levels allowable for a given operating point. The ODE reproduces the dominant behavior across most cases and provides an interpretable screening tool, but residual analysis reveals structured radius, impurity, and blanket-dependent bias, indicating that a single fixed closed-form expression is too rigid to fully capture all production and depletion patterns under one roof. This limitation is especially evident in specific impurity-blanket combinations, such as Ho in RAFM-Li4SiO4. To address this, we present a Kernel Ridge Regression (KRR) surrogate trained directly on operating and material descriptors, testing whether a more flexible data-driven model can better encode these patterns while preserving the rapid evaluation needed for impurity specification and design-space exploration.

Speaker: Jhovanna Garcia (ORNL)

KIT_Abstract_final.pdf

17:00 Comparative Analysis of Activated Corrosion Products in Key Systems of a Burning Plasma Experimental Tokamak

Activated corrosion products (ACPs) are a key source term for shutdown dose rates and radioactive waste management in fusion reactors. In this paper, we focus on a burning plasma experimental tokamak and apply the CATE (Corrosion, Activation and Transport Evaluation) code to analyze four key systems: blanket first wall, shield block, vacuum vessel, and divertor. We calculate the evolution of ACPs during pulsed operation (0–10 years) and the post-shutdown period (10–40 years), examine the influence of coolant water chemistry, and track changes in main radionuclides. Our results show that immediately after shutdown, the blanket first wall exhibits the highest activity, followed by the shield block and the divertor, while the vacuum vessel shows the lowest. During the post-shutdown period, activity in all systems decays over time, but the blanket first wall remains the highest throughout. Increasing the coolant pH generally reduces the production of ACPs. The main radionuclides in the coolant evolve in distinct stages: during early operation, Cu-64, Cr-51, and Fe-55 prevail; at later cooling times, long lived nuclides such as Ni-63, Fe-55, and Co-60 become dominant. This work provides a quantitative basis for source term control and personnel protection in fusion reactors.

Speaker: Chen Yang (Institute of Energy, Hefei Comprehensive National Science Center)

Abstract.docx

17:00 Differentiable deterministic neutronics for stellarator design

Neutronic engineering constraints significantly limit the stellarator design space by forcing a minimum coil-plasma distance compliant with sufficient tritium breeding ratio and fast flux shielding. Currently, these engineering constraints are taken into account by simply posing a minimum coil-plasma distance, independent of the specific coil or blanket design. Afterwards, a blanket is designed and checked whether this satisfies the constraints. This is necessarily an iterative process and may not arrive at an optimal solution. Including the actual neutronics constraints in the plasma-coil optimization process could therefore reduce iteration times and improve the performance of the total plasma-blanket-coil system. However, conventional Monte-Carlo based methods are difficult to integrate in optimization loops due to the required derivative computations. The already significant cost of a 3D stellarator simulation (required for a localized fast flux measure) is further amplified by the need for inefficient finite differences, leading to insurmountable computational costs.

In previous work, we have developed [1] and benchmarked [2] a deterministic neutronics model. In principle, deterministic neutronics models can provide significant speedups at the cost of accuracy and memory. More importantly, although it requires significant implementation effort, adjoint-based algorithmic differentiation can provide the required derivatives at the cost of only one extra solve (irrespective of the number of optimization variables). Results from the previously published benchmarks show that this deterministic neutronics model can calculate the neutronic engineering constraints within an hour while retaining acceptable accuracy (<10% difference between the deterministic model and OpenMC). It also generates high-quality weight windows using the FW-CADIS and CADIS methods.

In this work, we extend the deterministic model with GPU acceleration and the adjoint-based derivative computation using a combined JAX and NVIDIA Warp implementation. The GPU acceleration further speeds up the method, such that results are obtained within 2 minutes Furthermore, the parametric geometry workflow has also been rewritten to support algorithmic differentiation, such that the entire stellarator design pipeline can be differentiated. As proof of concept, we will show several applications, ranging from 1D multi-objective optimization to the optimization of a layered, 3D, stellarator blanket that minimizes the required coil-plasma distance while respecting the engineering constraints.

[1] Bogaarts, T. J., & Warmer, F. (2025). A novel discontinuous-Galerkin deterministic neutronics model for fusion applications: development and benchmarking. Nuclear Fusion, 65(7), 076015.
[2] Bogaarts, T. J., & Warmer, F. (2026). A novel discontinuous-Galerkin deterministicneutronics model for fusion applications: workflowfor stellarator reactor design studies. Nuclear Fusion.

Speaker: Timo Bogaarts (Eindhoven University of Technology)

abstract_tj_bogaarts_f_warmer.docx

17:00 Isomeric Branching and GENDF XS in OpenMC Deplete

Full treatment of isomeric branching is important in predicting activation in high yield fusion generators. Here-to, OpenMC1 Deplete lacked energy dependent isomeric branching. Rather, pre-processed fixed branching ratios (BR) were stored for reactions on the chain. This work will present implementation of full energy-dependent isomeric branching in OpenMC, accomplished by leveraging UKAEA GENDF2 activation libraries. These libraries are ENDF evaluations processed into group-averaged form by PREPRO, with partial cross-sections (XS) to each isomeric product consolidated on MF=10. The implementation is ultimately applied to Helion's currently operational fusion prototype, Polaris.

First an augmented OpenMC chain is produced, reactions with isomeric branching flagged and GENDF excited states matched with decay metastable states. During deplete, multi-group flux with GENDF energy structure is tallied. When a pathway with a BR flag is encountered a helper fetches the GENDF MF=10 XS. The flux and partial XS are collapsed, forming BRs. These are applied to the reaction rate, unfolding it into multi-target isomeric pathways.

With this GENDF groundwork, further enhancements were made beyond BR.
1. GENDF XS as the activation XS, in-lieu of the continuous energy transport XS. Benefits:
a. wider nuclide coverage (especially metastables)
b. faster transport solve and XS collapse (albeit coarser energy structure)
2. Addition of MT=4 (n,n’) neutron inelastic activation. Previously absent without the requisite pipework for isomeric pathways e.g. no In115(n,n’)In115m products.

CoNDERC FNS Decay-Heat Benchmark3 results will be presented. Improvements are stark:
a. significantly improved agreement with experiment
b. near perfect agreement with FISPACT-II2

Speaker: Perry Young

Young_Abstract_FNM2026_IsoBR-GENDF_OMC.pdf

17:00 MCNP Simulation of gamma background in the ITER Radial Gamma Ray Spectrometer

Collimated line-of-sight diagnostics such as the Radial Gamma Ray Spectrometer (RGRS) could be used for fusion power measurement at ITER. Compared with other approaches, such as fission chambers or activation foils, they offer the advantage of observing only the plasma section along a collimated line of sight, simplifying the calibration procedure by removing the need for complex neutronic calculations, such as the calculation of the adjoint flux for every point of the plasma. However, calculating neutron and gamma fluxes in regions located tens of meters from the plasma presents the unique computational challenge of low particle statistics.
This work presents MCNP-based Monte Carlo simulations aimed at characterizing the gamma background expected at the RGRS LaBr₃ detectors during full-power ITER operation. The ITER C-model was updated to reflect the latest RGRS design changes, including new collimator diameters, reoriented detectors aligned with the lines of sight, and an updated representation of Radial Neutron Camera. A two-step simulation method was developed to estimate the contribution of neutron-induced prompt gamma radiation originating from the central column, that was identified as the primary source of gamma background along the line of sight, contributing to a gamma current roughly one order of magnitude higher than that from all other sources combined.
The expected detector count rate under full-power conditions was estimated at approximately 1.4x107 counts per second. Model alignment was verified by comparing transport factors between MCNP and the Linalytic code, yielding agreement within 2.5%. These results provide an improved basis for the performance evaluation of RGRS and inform future background studies, including contributions from nearby lines of sight, such as that of the High-Resolution Neutron Spectrometer (HRNS).

Speaker: Stefano Colombi (University of Milano-Bicocca)

abstract_v4.docx

17:00 Neutron flux characterisation inside the inner irradiation Snail head of the KATANA facility

KATANA is a water activation facility installed at the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. It was developed to investigate water activation processes relevant to future fusion systems such as ITER. The primary objective of KATANA is to perform benchmark-quality experiments, e.g. the validation of the state-of-the-art fluid activation codes, for which an accurate characterisation of the neutron flux distribution in the irradiation region is essential. The neutron flux directly determines reaction rates and the resulting activity of the main radioactive isotopes (N-16, N-17, and O-19) throughout the water circuit. This work presents an experimental analysis of the neutron flux inside the inner irradiation “Snail head” of the KATANA facility, located in close proximity to the reactor core, where most of the water activation occurs.
Measurements were performed using the Libera MONACO 3 digital acquisition system and two miniature fission chambers (FC), with fissionable coatings primarily composed of U-235 and U-238. These provide sensitivity predominantly to thermal and fast neutrons, respectively. The characterisation included investigations of the detector response as a function of reactor power to establish appropriate operating regimes for counting and current modes, as well as measurements at different radial positions along the Snail head.
The measured relative radial fission rate distributions for the U-235(n,f) reaction (FC U-235) show good agreement with MCNP calculations, thereby validating both the computational model of the reactor/KATANA system and the detector performance. In contrast, significantly larger discrepancies were observed for the FC U-238: the experimental results exhibit a trend similar to that of the U-235 chamber rather than the expected U-238(n,f) response. This indicates that fast neutrons are not being detected properly, likely due to a significant contribution from thermal neutrons, and that further optimisation is required. Proposed improvements include additional coating or shielding against thermal neutrons and/or the use of a FC with ultra-high-purity U-238 coating (99.9999 %). Future work will also focus on absolute neutron flux determination through neutron activation analysis using a dedicated set of activation foils.

Speaker: Dr Domen Kotnik (Jozef Stefan Institute)

10_04_202626_Kotnik_Domen_FNM_V1.docx

17:00 Neutronic analyses for the shielding design of the DTT Neutral Beam Injector Transmission Line

The Divertor Tokamak Test (DTT) is a new experimental facility designed to the study of plasma exhaust issues in a DEMO-relevant environment and to test different divertor configurations. During high-performance operation, DTT is expected to produce approximately 1.5e17 neutrons per second from D-D fusion reactions. In addition to that a non-negligible amount of neutrons produced from D-T reactions are also expected, due to the triton burn up. The building that will host the DTT device must provide adequate radiation shielding to protect both workers and the public from neutron and gamma radiation.
A three-dimensional (3D) neutronic analysis has been performed to evaluate the effective dose rates in areas outside the main penetrations of the north wall of the Torus Hall Building (THB). Among these cut-outs, the Neutral Beam Injector (NBI) Transmission Line (TL) features an opening of about 2.3 m in diameter is foreseen, through which a significant amount of neutron and gamma radiation escape from the THB, causing an increase of the effective dose rate to the workers and the public.
This work therefore focuses on assessing the neutron and gamma streaming through the NBI TL penetration, with the aim of optimizing and evaluating the effectiveness of several proposed shielding solutions to ensure adequate worker protection in terms of occupational dose.
Neutron and gamma transport simulations were carried out with the MCNP5 Monte Carlo code employing variance reduction techniques (i.e. weight windows) generated with the Global Variance Reduction (GVR) tool.
Results of the analyses are presented and reported along with some considerations to improve the shielding capabilities.

Speaker: Samuele Castegnaro (ENEA)

Abstract_template_Samuele_Castegnaro.pdf

17:00 Neutronic and Structural Optimization of a Novel Pin-Type HCPB Breeding Blanket for the DEMO LAR

This study introduces a novel Helium Cooled Pebble Bed (HCPB) Breeding Blanket (BB) design tailored for the DEMO Low Aspect Ratio (LAR) reactor. Despite the stringent constraints imposed by reduced fusion power and compact tokamak dimensions, the proposed design maintains the conventional pin-type architecture while achieving a sufficient Tritium Breeding Ratio (TBR) and superior internal shielding performance. Crucially, this configuration significantly mitigates the shielding deficit typically associated with helium-cooled systems compared to water-cooled counterparts, while simultaneously addressing the fabrication challenges related to welding in traditional pin structures. Structural modifications within the Breeder Zone (BZ) have notably revitalized the effectiveness of shielding blocks in the Back Supporting Structure (BSS) a component that previously offered negligible benefits. Consequently, the enhanced radiative shielding capability of the new HCPB design ensures substantial safety margins for the Vacuum Vessel (VV) and Toroidal Field Coil (TFC), thereby satisfying rigorous design safety factors. Finally, to further refine the design, neutronics simulations were conducted with various candidate shielding materials applied to the BSS shielding block, identifying the optimized material composition and design parameters.

Speaker: Jin Hun Park (Karlsruhe Institute of Technology)

17:00 Nuclear Analysis of the High Resolution Neutron Spectrometer (HRNS): Activation and Radioactive Waste Assessment

The High Resolution Neutron Spectrometer (HRNS) is designed for neutron diagnostics in ITER, where accurate measurements must be combined with careful assessment of radiation-induced activation of detector and structural components. Understanding the activation behaviour of diagnostic systems is essential for evaluating operational safety, maintenance requirements, and long-term radioactive waste generation.
This work presents a comprehensive nuclear analysis of the HRNS system, focusing on activation and radiological inventory of its main components. Neutron transport and activation calculations were performed for the full HRNS assembly, including shielding and structural components, detector systems (TPR, NDD, bToF and fToF), the intermediate collimator, the electrical cabinet, and the beam dump. The analysis investigates the time evolution of specific and total activity over cooling times from seconds to hundreds of years and identifies the dominant radionuclides governing the radiological behaviour.
The results show that the short-term activity of the HRNS system is dominated by structural stainless steel components located close to the neutron source. Austenitic steels containing nickel exhibit the highest specific activation, while large ferritic shielding components govern the overall radioactive inventory due to their substantial mass. The tungsten beam dump shows very high initial activation driven by medium-lived isotopes such as 185W and 181W. At longer cooling times, the residual activity becomes dominated by long-lived radionuclides in stainless steels (e.g. 55Fe, 60Co and 63Ni) and by tritium produced in boron-containing materials such as B₄C used in neutron collimation and absorption components.
The study demonstrates that the radiological behaviour of fusion neutron diagnostics is controlled by the combined effects of neutron exposure, material composition, and component mass. The results provide guidance for material selection and detector design aimed at minimizing activation and optimizing radiological performance of neutron diagnostic systems in future fusion facilities.

Speaker: Dr Anna Wójcik-Gargula (Institute of Nuclear Physics Polish Academy of Sciences)

Abstract_AWojcikGargula.pdf

17:00 OPERA: Advanced Radiation Transport Solution with CAD Integration and Hybrid Methodologies

OPERA, a CEA Paris-Saclay initiative, gathers cutting-edge tools for fixed-source radiation transport. In the early stages, our goals are:
• A 3D modeling tool, capable of CAD import and meshing;
• Enhanced hybrid capabilities, accelerating Monte Carlo fixed-source calculations with deterministic methods and offering standalone 3D deterministic calculations.

Speaker: Jean-Christophe Trama (CEA Paris-Saclay)

FNM_OPERA_submitted.pdf

17:00 Radioisotope Production in the Volumetric Neutron Source Tokamak

A variety of medical and radio isotopes yields have been calculated in Eurofusion's Volumetric Neutron Source (VNS) Tokamak using Serpent burnup mode simulations. The VNS is uniquely suited for the task of medical isotope production due to the fact that it will undergo a high neutron fluence requiring relatively low tritium (<1kg/year) that can be supplied by existing CANDU reactors. The VNS is designed to test tritium breeding blankets only, and does not aim to produce more tritium than it consumes, and therefore has additional irradiation volume available which can be used for medical isotope irradiation facilities. Such facilities have been modeled for isotope yield calculations under the fusion spectrum, as well as under a spectrum in which the capsules contain highly efficient Zirconium Hydride moderator.

Speaker: Christopher Ehrich (Forschungs-Neutronenquelle Heinz Maier- Leibnitz (FRM II) Technische Universität München)

17:00 Recent Developments in Fusion Neutronics Capabilities of the Reactor Monte Carlo code RMC

The Reactor Monte Carlo (RMC) code, developed by the Reactor Engineering Analysis Laboratory (REAL) at Tsinghua University, is a continuous-energy Monte Carlo transport code that has been widely applied in reactor physics analyses. In recent years, its capabilities have been further extended to fusion neutronics applications. This work summarizes recent progress in the development and validation of RMC for fusion neutronics analysis, with emphasis on model conversion, transport calculation, variance reduction, tritium breeding analysis, CAD-based modeling, and shutdown dose rate assessment.

An MCNP-to-RMC conversion tool, M2R, has been developed to support the transfer of existing fusion neutronics models into the RMC framework. Representative models, including the ITER C-model and the CFETR neutronics model, have been successfully converted and used in validation studies. Based on these models, neutron transport calculations have been performed to assess the applicability of RMC to fusion neutronics problems. For the CFETR model, global neutron flux distributions, variance reduction techniques, and Tritium Breeding Ratio calculations were investigated, with results showing good agreement with OpenMC. To further extend the application range of RMC in fusion analysis, its CAD-based transport capability has also been investigated using simple tokamak CAD models. Comparative studies with OpenMC showed consistent neutronics results, demonstrating that RMC can handle both conventional converted models and more flexible CAD-based geometries for fusion reactor analysis.

A further recent development is the establishment of a fully integrated Rigorous Two-Step (R2S) framework within RMC for shutdown dose rate calculations. By extending the existing criticality transport–depletion capability to a fixed-source transport–activation coupling framework and developing a decay-photon source generation module, RMC is able to perform neutron transport, activation calculation, decay-photon generation, and photon transport within a single code system. Verification against simplified OpenMC-based reference cases showed close agreement in nuclide inventories, decay-photon spectra, and dose rates. The framework was further applied to the ITER port plug benchmark for validation, where reliable shutdown dose rate results were obtained.

These developments show that RMC has made substantial progress as an integrated tool for fusion neutronics analysis, expanding from basic transport capability toward more comprehensive support for advanced fusion applications.

Speaker: Mr Kok Yue Chan (Department of Engineering Physics, Tsinghua University)

Recent Developments in Fusion Neutronics Capabilities of the Reactor Monte Carlo code RMC.pdf

17:00 W7-X neutronics simulations for new neutron diagnostics with application to fast-ion confinement

By the end of 2028 it is foreseen that the W7-X stellarator will switch from Hydrogen to Deuterium fuel, starting DD operation. The DD fusion reactions will lead to the production of 2.5 MeV neutrons together with fast Tritium nuclei which can themselves fuse with Deuterium thermal plasma creating 14 MeV DT neutrons.
The detection of these two different species of neutrons, namely 2.5 MeV DD neutrons and 14 MeV DT neutrons, is a well-established method, known as triton burn-up study, to assess the fast-ion confinement. Tritium burn-up studies were already employed in other fusion experiment devises, both tokamaks (such as ASDEX-Upgrade and JET), as well as stellarator (LHD). Understanding, improving and predicting the behaviour of fast ions, especially their confinement, in present-day fusion devices is a major scientific objective especially for stellarators.

Future fusion reactors will work on a DT fuel mixture whose reaction gives birth to highly energetic alpha-particles (3.5 MeV Helium-4 ions). For a burning plasma, the internal heating has to be provided by the alpha-particles, which must be well confined in the plasma to transfer their energy to the thermal plasma particles. Compared to less energetic particles, fast ions require different and more challenging conditions, as they are in a different collisional regime and less well confined by the magnetic field. Unlike in tokamaks, the three-dimensional magnetic field configuration of stellarators leads to losses of the helically trapped fast-ions. An optimization criterion for a good fast-ion confinement is a high plasma pressure. The validation and improvement of these conditions for the W7-X stellarator is a crucial step to demonstrate a possible path towards reactor relevant conditions in stellarators.

For a triton burn-up study, appropriate neutron diagnostics need to be set-up which can distinguish between fast neutrons of different energies such as DD and DT neutrons. Previous studies have researched the feasibility of using scintillating fiber (SciFi) neutron detector for W7-X triton burn-up measurement. These SciFi detectors are not installed yet at W7-X, therefore neutronics simulations are necessary to appropriately set up the diagnostics and interpret the detector response in relation to the physics. The aim of our research is to perform high-fidelity Monte Carlo neutronics simulations using OpenMC code on an updated CAD-based W7-X geometry model for the evaluation of the optimal positions of the SciFi detectors, their expected response and connection to fast-ion confinement analysis. In broader terms, this study will contribute to the development of a simulation technique which can be applied to neutronics analyses for neutron diagnostics evaluation in future fusion reactors.

Speaker: Riccardo Brambilla (TU Eindhoven, IPP Max Planck)

Abstract.pdf

Tuesday 9 June

09:00 → 10:00 Invited Session

09:00 Is It Time for a Fusion-Specific Monte Carlo Code?

Fusion neutronics has long relied on tools built for fission, adapting transport codes designed around criticality calculations, borrowing finite-element meshers intended for structural analysis, and working around defaults tuned for reactor physics. With a growing fusion industry, increasing demand from private companies, and the reduction in development effort enabled by large language models, it is worth asking: has fusion become large enough to justify its own purpose-built neutronics workflow?

The 14.1 MeV source neutrons in fusion open reaction channels that fission neutrons lack the energy to reach, resulting in activation analysis across a different range of progeny nuclides. Combined with complex streaming path geometries and pulsed operation, fusion workflows present distinct challenges compared to fission. General-purpose codes tend to ship with defaults tuned for fission, from the reactions included in transmutation chains to normalisation factors in post-processing. These are common sources of user error and additional effort in fusion workflows.

Many Monte Carlo transport codes are closed-source or export-controlled (e.g. MCNP). Existing open-source options such as OpenMC have made great progress in accessibility and licensing. However, CAD-based transport workflows that are critical for fusion still rely on external meshing tools that often carry copyleft licenses, which the growing fusion private sector frequently restricts.

Stellarators such as the Proxima Fusion designs, with their non-axisymmetric, twisted geometries, make this especially acute as their plasma-facing components and coil structures simply cannot be faithfully represented in CSG. Fusion devices arguably demand CAD or mesh-based representations as a first-class capability. A fusion-specific code could also include a mesher optimised for particle transport rather than relying on finite-element meshers designed for a different problem.

Narrowing scope to fixed-source transport could bring further advantages: a smaller, more maintainable codebase, less likely to be export controlled, free of copyleft meshing dependencies, and potentially a more natural fit for GPU acceleration. Were such a code started today, AI-accelerated development and modern packaging could significantly reduce the effort involved.

There are clear arguments against, too. Building and validating a new transport code is a significant undertaking, fewer fusion-relevant benchmarks exist compared to fission, and a new code would need to earn trust that established codes have built over decades. Fragmenting developer effort and losing cross-pollination between the fission and fusion communities are real risks.

This work explores what a purpose-built fusion Monte Carlo code could look like and if the perceived benefits justify the endeavour.

Speaker: Dr Jonathan Shimwell (Proxima fusion)

fusion-specific-MC-abstract-kit-neutronics-meeting.pdf

09:30 FERMI: Fusion Energy Reactor Models Integrator

Speaker: Dr Vittorio Badalassi (ORNL)

10:00 → 11:00 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

10:00 An overview of STEP Breeder Blanket nuclear analysis using OpenMC

The Spherical Tokamak for Energy Production (STEP) is a UK programme to design and build a prototype fusion energy plant. STEP is an ambitious programme that will demonstrate the ability to generate net energy, fuel self-sufficiency and a route to commercialisation of nuclear fusion.

The compact radial design of a spherical tokamak presents a significant neutronic design challenge for tritium production that meets self-sufficiency targets, with inboard breeding challenging to implement. Following a comprehensive assessment of all breeder, coolant, and structural material options, solid ceramic lithium oxide (Li2O) has been selected together with a Ti-modified austenitic stainless steel structural material, CO2 coolant, and beryllium-based multiplier. This combination is considered to give the highest confidence in successfully meeting the SPP requirements.

Nuclear performance such as TBR, nuclear heating and shielding were key factors in the choice of blanket architecture for STEP Prototype Powerplant (SPP) breeder blanket using these materials. This talk presents the workflow used to perform nuclear analysis on various breeder designs and the challenges of an optimization process that must be integrated with design, thermal and structural analysis.

A Constructive Solid Geometry (CSG) reactor model incorporating parameterised CAD in DAGMC universes was used to rapidly explore the design space of high-fidelity blanket architectures, allowing quantitative comparison between configurations. The selection of an architecture and progression from concept to detailed design has led to an increase in design complexity, requiring improved workflows that can incorporate key features of the design for timely neutronics analysis to support integrated design choices. This talk presents some of the updates to the workflow required to manage this complexity, and plans for further improvement.

This work has been funded by STEP Fusion, a major technology and infrastructure programme led by UK Industrial Fusion Solutions Ltd (UKIFS). UKIFS will shortly be renamed to UK Fusion Energy Ltd, reflecting its role in the next phase of the national fusion mission.

Speaker: Henry Marden (UK Atomic Energy Authority)

Abstract_FNM2026_blankets_v2.pdf

10:20 Parametric Multiphysics Workflow for Evaluating Tritium Breeder Blanket Performance

The efficient and accurate design of tritium breeder blankets is essential for the success of nuclear fusion reactors, playing a vital role in achieving optimal performance and safety. This study, developed within the LIBRTI programme funded by UK Atomic Energy Authority (UKAEA), introduces a comprehensive workflow that integrates parametric geometry generation, meshing, and multiphysics simulation to analyze tritium breeder blanket architectures, with particular emphasis on the methodology for correlating scaled mock-up results to full-size blanket performance.

The workflow begins with parametric geometry modeling using ParaBlank, an open-source Python tool developed by IDOM for the STEP programme. ParaBlank integrates CadQuery-based parametric geometry generation, high-quality conformal meshing with Gmsh, and geometry conversion to DAGMC for neutronics. Material properties, boundary conditions, and physical parameters are embedded directly during geometry creation, ensuring geometric consistency across disciplines by preserving face-level tags.

The SALAMANDER platform, developed by Idaho National Laboratory, serves as the core simulation framework, integrating key physics domains: neutronics via OpenMC, thermal-hydraulics and thermomechanics via MOOSE, and multiscale tritium transport via TMAP8 for predicting tritium release to the purge gas.

A key methodological contribution addresses the computational challenge of full-scale blanket simulation. At the mock-up scale, detailed multiphysics simulations produce data for global sensitivity analysis to identify parameters governing blanket performance. These parameters inform surrogate models that represent detailed pin-level physics at reduced computational cost. At full-size blanket scale, where detailed 3D simulations become computationally impractical, system-level analyses leverage these surrogate models combined with the MOOSE Thermal Hydraulics Module to efficiently capture essential local physics within each breeder unit.

The analysis focuses on critical performance metrics including neutronic heat deposition, Tritium Breeding Ratio, thermal performance, structural integrity, coolant behavior, and tritium release characteristics. The derived correlations between mock-up and full-scale behavior provide insights for experimental validation and their implications at fusion reactor level.

Speaker: Alexandre Sureda Croguennoc (IDOM)

Abstract_Alexandre_Sureda_Croguennoc_IDOM.docx

10:40 Neutronics Analysis of Ceramic Breeder blankets for the Infinity 2 Stellarator

Infinity Two is a 350 MWe fusion pilot power plant under development by Type One Energy, designed to be one of the first commercially viable stellarator based fusion energy systems. A central feasibility challenge for this advanced high field stellarator concept is the development of an efficient and fully integrated blanket system. Consequently, the design and optimisation of the breeding blanket form a critical component of the Infinity Two physics basis.
This work presents the application of OpenMC, combined with CAD derived geometries, to optimise a ceramic breeder blanket for maximised tritium production and heat generation. Unstructured mesh techniques are employed to resolve spatial distributions of key nuclear performance metrics across complex blanket geometries. Parametric studies investigate the influence of enriched material fractions, packing density, multiplier material quantity and placement, and overall blanket topology. Across the design space explored, multiple promising configurations are identified, achieving tritium breeding ratios (TBR) greater than 1.25 along with sufficient energy multiplication to support Infinity Two’s 350 MWe power target.

Speaker: Jonathan Naish (Type One Energy)

2nd_Fusion_Neutronics_Meeting_2026_Abstract.docx

11:00 → 11:30 Coffee Break

11:30 → 12:30 Neutron Source Facilities Design and Exploitation

Performance evaluation
Neutron/gamma diagnostics and calibration

11:30 Predictive neutron source modelling for the ITER 15MA/5.3T D-T baseline scenario

In the contribution we present a framework for computing realistic plasma neutron source models for ITER, based on coupling between the plasma transport code JINTRAC and the fusion kinematics code DRESS, and demonstrate its application to the MCNP neutron transport code. This allows capturing plasma events throughout the whole predicted scenario trajectory, including transients, and their direct effect on neutron emissivity properties and nuclear detector response. The ITER 15MA/5.3T D-T baseline scenario is modelled throughout the beginning of the flat-top operational point, in which n/(nD+nT), Pfus and Qfus are ramped-up from negligible levels to steady-state operation values at ≈ 0.5, 500 MW and 10, respectively. Four trajectory points within the fusion power ramp-up phase are modelled, with different plasma heating configurations and D-T composition ratios. For each snapshot the plasma state is modelled with state-of-the-art core-edge-SOL coupled JINTRAC simulations in IMAS, and the distribution of energetic beams ions calculated with ASCOT4. Large differences between the core and edge D-T ratio are found at the beginning of the ramp-up phase, affecting fusion reactivity profiles and limiting the operational domain of nuclear diagnostics, which arise due to an interplay of D gas puffing, D-T pellet fueling, and inward fuel transport. DRESS-computed emissivity profiles and neutron energy spectra are compiled into a bespoke MCNP SDEF source and launched within the ITER C-lite neutronics model. A neutron detector sensitivity is presented to address the significant differences in shape and spectra observed between the standard and new ITER source models. The JINTRAC-DRESS-MCNP workflow is being made compatible with the IMAS infrastructure to facilitate a seamless exchange of data in IDS format and routine creation of neutron source models for arbitrary ITER scenario plasma simulations, aligned with the new research plan.

Speaker: Žiga Štancar (UK Atomic Energy Authority)

Štancar_ITER-Predictive-Source-Model_Abstract-FNM_v2.pdf

11:50 Shielding Optimization for a Neutron Source Using Monte Carlo Methods

This work presents a systematic optimization study of neutron source shielding for a dedicated multilayer bunker designed to house a cylindrical inertial electrostatic confinement (IEC) neutron source within the framework of the LIBRA project. LIBRA aims to investigate lithium deuteride (⁶LiD) as a tritium breeding material for future nuclear fusion applications.

The bunker design must guarantee dose rates below 2 μSv/h at 1 m from the external wall to limit radiation exposure to personnel. The optimization process is carried out using the Monte Carlo neutron transport code OpenMC. A parametric sweep methodology is employed to explore a wide range of shielding configurations, systematically varying material selection and layer thicknesses. The methodology begins with the evaluation of individual shielding materials, followed by a comprehensive parametric exploration of the complete bunker geometry that considers all relevant material and thickness combinations and includes skyshine contributions. The analysis is completed by evaluating the impact of additional radiation-mitigation features, such as absorbing panels and the access doors.

The results show the influence of individual materials and thicknesses to be quantified, revealing key trends and their impact on dose reduction and overall shielding performance. The suggested methodology provides a robust and flexible framework for shielding design in fusion-related neutron facilities and can be readily extended to similar experimental platforms.

Speaker: Sara Abad Jiménez (IDOM)

SARA_ABAD_KIT.pdf

12:10 Evaluation of Shutdown Dose Rate at JET after the second Deuterium–Tritium Campaign

Accurate evaluation of shutdown dose rate (SDDR) represents a key requirement in fusion reactor design and operation and is essential to ensure personnel safety throughout the entire lifetime of fusion facilities, in particular during maintenance and their decommissioning. SDDR tools have been developed for this purpose, among which JSIR2S, developed at the Jožef Stefan Institute, implements the rigorous two-step (R2S) methodology. In this approach, neutron transport simulations are first carried out using the Monte Carlo transport code MCNP, followed by activation and inventory calculations of irradiated materials with FISPACT code, to produce the gamma source of the subsequent MCNP run. In the second step, photon transport calculations are performed to determine the resulting dose rates from radioactive decay.
The JSIR2S code has previously been benchmarked against experimental data from the TRIGA research reactor, and the present study extends its validation to fusion-relevant conditions using SDDR measurements after the second deuterium-tritium campaign (DTE2) at JET. The calculated SDDR is compared against some selected experimental measurement points and against the predictions of the other computational tools involved in the same benchmark campaign, thus enabling a comprehensive assessment and critical comparison of the methodology and of the specific tool. SDDR calculations at JET following DTE2 are presented in this work with the primary objective of validating the computational predictions of JSIR2S in a fusion-relevant environment as JET.
The comparison shows good agreement between calculated and measured values over the considered cooling times, up to 29 days after shutdown. For the last set of pulses of the campaign, C/E values for ionisation chambers in Octant 1 range between (0.75 ± 0.08) and (1.36 ± 0.07). In Octant 2, lower agreement is observed, with C/E values around (0.11 ± 0.01). For Octant 1, the C/E values are comparable to those obtained with the Advanced D1S method, whereas for Octant 2 the Advanced D1S shows better agreement with experimental data.

These results indicate that the applied approach with the JSIR2S code provides generally reliable predictions of SDDR in fusion environments, although the observed discrepancies warrant further investigation.

Speaker: Ylenia Kogovšek Žiber (Jožef Stefan Institute (JSI))

Abstract_template_YKZ_v1.docx

12:30 → 13:00 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

12:30 FENDL: Status and Future

The Fusion Evaluated Nuclear Data Library (FENDL) has been serving the Fusion Neutronics community for more than three decades, particularly addressing the nuclear data needs of the ITER project, for which it is the design library since FENDL 2.1, released in 2006.
Its development is driven by an international committee and coordinated by a series of meetings under the auspices of the IAEA Nuclear Data Section (NDS). Over the course of time, FENDL has been significantly updated to meet emerging data needs, such as an increase in isotope coverage and an extension of incident energies, which have been carried out in an IAEA Coordinated Research Project from 2008 to 2012.
More recently, a significant focus has been put on making the library development more transparent and traceable, starting from the compilation of source nuclear data files (in ENDF format) over processing with NJOY2016 and ending with streamlined verification and validation (V&V) with the JADE software package.
A comprehensive milestone paper for FENDL-3.2b has been published recently, which contains history, data preparation and processing, library content and V&V results [1]. A follow-up paper provides documentation of FENDL-3.2c, a maintenance release featuring improved processing by NJOY2016 with patches developed and tracked by the IAEA NDS [2].
This contribution will summarize the status of the FENDL project and present recent, ongoing, and envisioned developments, with a focus on robust methods and processes to enhance efficiency and quality assurance in the preparation of the next version of FENDL.

[1] G. Schnabel, D.L. Aldama et al. Nuclear Data Sheets 193, 1-78 (2024)
[2] T. Bohm, D.L. Aldama, et al. Fusion Science and Technology, 1-15 (2025)

Speaker: Georg Schnabel (IAEA)

Abstract - FENDL - status and future.docx

13:00 → 14:00 Lunch Time

14:00 → 15:20 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

14:00 A Reliability-Based Framework for Prioritizing Nuclear Data Needs: Application to the Fusion Nuclear Science Facility

Recent community discussions, including WANDA 2024, have highlighted the need for sensitivity studies to prioritize nuclear data improvement efforts for Fusion Energy Sciences. Yet nuclear data sensitivity analyses in fusion are still commonly performed one response at a time, such as tritium breeding ratio or activation, and one component at a time, such as the blanket, first wall, or vacuum vessel, producing competing prioritization lists of important nuclides and reactions. Because fusion power plants are tightly coupled systems that must satisfy multiple constraints simultaneously, a whole-reactor methodology is needed to connect nuclear data needs directly to reactor feasibility and reliability.

Here, we establish a systematic framework for prioritizing nuclear data needs in fusion energy by quantifying how nuclear data uncertainties affect the probability that a power plant design fails to meet key performance requirements. The central idea is to replace isolated, response-by-response prioritization with a reliability-based approach that integrates multiple reactor criteria into a single performance measure. Using Monte Carlo neutron transport with nuclear data uncertainty propagation, together with activation and cost-relevant modeling, the methodology identifies the nuclides and reaction channels that most strongly influence overall reactor viability.

As a first step, the framework is applied to a simplified 1D cylindrical radial build model of the Fusion Energy Systems Study Fusion Nuclear Science Facility (FESS–FNSF), derived from a detailed 3D reference design. The model represents the radial material layout with 85 zones and captures major blanket and shielding features in the inboard and outboard regions, including PbLi flow channels, SiC flow-channel inserts, and helium-cooled structural components. Nuclear data uncertainties are propagated by processing libraries with NJOY, sampling evaluated data with SANDY, and calculating neutronics responses with OpenMC.

The resulting workflow quantifies uncertainty in key responses and provides a baseline for subsequent sensitivity analysis and nuclear data prioritization. It is designed to extend across multiple fidelities, from 1D radial builds to conceptual and detailed 3D reactor models, ultimately enabling a whole-reactor view of nuclear data needs that is usable by both public-sector programs and private fusion enterprises. The long-term outcome is an open-source framework to guide high-impact, cost-effective nuclear data improvement for fusion energy development.

Speaker: Enrique Miralles-Dolz (Princeton Plasma Physics Laboratory)

emd_KIT_Neutronics_Abstract.pdf

14:20 Latest V&V results on FENDL 3.2c and OpenMC using JADE

JADE is a python framework that makes it easy to automatically pre-process, run and post-process large numbers of neutronic simulations which are used for the Verification and Validation (V&V) of nuclear data libraries and Monte Carlo codes.

Last year, at the first edition of the “Fusion Neutronics Meeting”, the new architecture of JADE v4 was presented. The complete rewriting of the software greatly simplifies the addition and maintenance of benchmarks that are part of the JADE suite and allows a full native integration of OpenMC into the framework.

During this last year, the project entered a phase of consolidation which focused on the quality of the post-processing, resolution of bugs and quality-of-life improvements for JADE users. The latest version of JADE (v4.4.0) was employed to perform a deep V&V assessment of the FENDL 3.2c (new recommended library for ITER project) using both MCNP and OpenMC. The work presents the main differences found between FENDL v3.2c and other major libraries as well as differences between MCNP and OpenMC results.

Speaker: Davide Laghi (Fusion For Energy)

14:40 Evaluation of Nuclear Data Libraries and Simulation Methods for IFMIF-DONES Test System

Safe licensing and operation of future fusion reactions requires demonstrating material performance after extreme irradiation conditions. This needs to be demonstrated through material irradiation in a suitable neutron source facility under fusion-relevant conditions

IFMIF-DONES is an accelerator-based neutron irradiation facility, providing the necessary irradiation data for the qualification of materials for the DEMO fusion power plant. This is achieved by the interaction of a 125 mA, 40 MeV deuteron beam with a liquid lithium courting, producing up to 6.75x1016 n/s. Additional secondary radiation sources arise from interactions between the deuterons and interceptive components such as scrapers and collimators used for beam shaping, as well as beam dumps and the beam duct. These interactions generate activation induced by both deuterons and secondary neutrons. The activation will give rise to delayed radiation fields during the shutdown periods, when the maintenance activities are carried out. These radiation fields need to be characterized to minimize the risk of exposure during both the operation and shutdown of the machine.

In this work, we first investigate the differences in the neutron source in IFMIF-DONES due to Li(d,xn) reactions when using different nuclear data libraries and simulation approaches. We also compared the results of simulations using different nuclear data libraries and simulation approaches with available experimental information, in order to obtain a better understanding of the application to IFMIF-DONES. In general, a good agreement between methodologies was found.

Finally, we also investigated the activation process and the applicability of the Direct-One-Step (D1S) method to IFMIF-DONES. The D1S methodology allows obtaining responses associated with decay photons in a single simulation thanks to the coupling of transport and activation calculations. But this approach relies on the assumption that radioisotopes are produced by one-step reactions, as well as negligible radioisotope burnup (Single Neutron Interaction and Low Burnup criteria). We identified the most relevant activation chains due to the neutron and deuteron irradiation, considering all the possible exposed materials, as well as typical spectrums, irradiation and decay times. Only two reaction chains with more than one step were found, verifying the D1S applicability.

Speaker: Victor Lopez Ochoa (Universidad Nacional de Educación a Distancia (UNED))

Abstract_2FNM_VLO_V2.docx

15:00 Sensitivity Analysis and Uncertainty Quantification of Nuclear Data for Tritium Breeding Ratio in Fusion Reactor Blankets Based on the Differential Operator Perturbation Method

In the design of fusion blanket systems, the Tritium Breeding Ratio (TBR) is a key parameter that characterizes the blanket's ability to sustain tritium. The accuracy of its calculation largely depends on the reliability of nuclear data and computational methods. Based on the Monte Carlo software NECP-MCX, this study uses the differential operator perturbation method to conduct a sensitivity analysis of nuclear data on TBR, focusing on the blanket of the China Fusion Engineering Test Reactor (CFETR). In addition, the nuclear data covariance matrix is incorporated, and the sandwich theorem is applied to quantify the uncertainties in the nuclear data. The results show that different nuclides and reaction channels have significant differences in their impact on TBR. Among all the considered nuclides, the relative sensitivity coefficients of reactions such as Be9 (n, 2n), Li6 (n,t), Be9 (n, nel), and Fe56 (n, nel) make a significant contribution. Further uncertainty analysis indicates that nuclides such as Li6 and Be9 are the primary sources of nuclear data uncertainty in TBR.

Speaker: QingMing He

Sensitivity Analysis and Uncertainty Quantification of Nuclear Data for Tritium Breeding Ratio in Fusion Reactor Blankets Based on the Differential Operator Perturbation Method.pdf

15:20 → 15:50 Coffee Break

15:50 → 16:50 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

15:50 Concrete Shielding Experiment at FNG: Benchmarking MCNP Simulations using JEFF 3.3 and JEFF 4.0 nuclear data libraries

Concrete is a widely used material in nuclear facilities, combining structural mechanical performance with effective radiation shielding capability. The design of bioshields and reactor buildings strongly relies on the ability of computational tools to accurately predict neutron and photon transport. In this context, the validation of numerical codes against experimental benchmarks is essential to ensure the reliability of safety assessments and shielding design. In recent years, the EUROfusion consortium has supported experimental campaigns to investigate neutron transport in concrete under fusion-relevant irradiation conditions: this work presents the results of the concrete shielding experiment, carried out at the Frascati Neutron Generator (FNG) facility at the ENEA Frascati Research Centre, using a 14 MeV neutron source.
The experimental setup consisted of a concrete mock-up assembled from 13 slabs (50×50×10 cm3 each), irradiated with 14 MeV D-T neutron source. Activation foils (Al, Au, Fe, In, Nb, Ni, W) were placed at specific positions within the assembly, providing sensitivity to a broad neutron energy range, through reactions characterized by different energy thresholds. Gamma spectroscopy after each irradiation was performed with corrections applied for dead time, self-absorption, geometric efficiency, and decay during both irradiation and acquisition. A detailed post-analysis has been carried out with a high-fidelity MCNP5 model of the full experimental setup, including the irradiation facility and surrounding bunker, to calculate reaction rates and determine calculated-to-experimental (C/E) ratios. A comparison of the results using JEFF 3.3 and JEFF 4.0 nuclear data libraries was performed.
Results confirm a generally reliable description of fast neutron transport through concrete. High-threshold reactions show good agreement: ⁹³Nb(n,2n) slightly underestimates measurements (~10%), while ⁵⁶Fe(n,p) closely reproduces experimental data. Larger discrepancies are observed for ²⁷Al(n,α) and ⁵⁸Ni(n,p). ¹¹⁵In(n,n') shows a systematic positional dependence, indicating sensitivity to local variations of the neutron energy spectrum within the mock-up. Thermal reactions display greater deviations, consistent with their increased sensitivity to neutron moderation effects. Generally JEFF 4.0 improves agreement relative to JEFF 3.3, particularly for thermal reactions, likely due to updated cross-section evaluations. Nevertheless, residual discrepancies, even among reactions with similar energy thresholds, highlight effects beyond the transport modelling, emphasizing the need for further sensitivity and uncertainty analyses.

Speaker: Virginie Lombardi (DIEE Department, La Sapienza University of Rome, 00186, Roma, Italy)

Abstract_Lombardi.pdf

16:10 Overview of measurements with silicon-carbide, lithium glass and optical fibers detectors in the FNG Concrete Shielding Experiment

In 2025, the EUROfusion concrete shielding experiment was conducted at the 14 MeV Frascati Neutron Generator (FNG) facility at ENEA Frascati with the aim of investigating neutron transport in concrete and assessing the capability of Monte Carlo codes to reproduce experimental results.
Three types of active neutron detectors were employed: silicon carbide (SiC) detectors, GS20 Li-glass detectors, and silica-based optical fibres as an innovative solution for neutron detection. The experiment was designed to characterize the neutron field within a concrete mock-up consisting of thirteen slabs (50 × 50 × 5 cm³), where the detectors were placed at different depths to measure neutron flux attenuation.
Two SiC detectors were used simultaneously to measure fast and thermal neutrons. Fast neutron attenuation was determined through the inelastic ¹²C(n,α)⁹Be and ²⁸Si(n,α)²⁵Mg reactions in the bare detector, while thermal neutrons were measured using the same detector coated with a thin ⁶LiF layer. Additionally, four identical ⁶Li-enriched Li-glass detectors were employed in a separate campaign to assess thermal neutron measurements. Both detector types were deployed at eleven identical measurement positions.
Furthermore, three 1 m long Ce-doped optical fibres, including one co-doped with Gd, were placed at seven positions to evaluate their capability to measure the neutron flux at different penetration depths.
The results are presented through a comparison of attenuation curves obtained from all detectors. Together with activation foil measurements, this analysis provides improved insight into neutron behaviour in concrete, which is a key shielding material in nuclear applications.

Speaker: Andrea Colangeli (ENEA)

FNM_abstract_Colangeli_final.docx

16:30 The ICSBEP Concrete Labyrinth Benchmark: A comparative analysis between OpenMC and MCNP

A comparative assessment between OpenMC 0.15.3 and MCNP 6.2 was performed on a three-section concrete labyrinth experiment from the ICSBEP database. The already existing MCNP benchmark model was converted to OpenMC and mesh-based weight windows were generated using the MAGIC method.

Neutron transport from a 252Cf spontaneous fission source was simulated, with neutron flux computed at multiple positions along the labyrinth. Results obtained with each transport code were compared and benchmarked against the available experimental measurements. The ENDF/B-VIII.0 nuclear data library was employed in both cases.

Multiple experimentally tested labyrinth configurations were simulated, including cases with added material to the labyrinth walls and polyethylene source filtering. Given the relevance of polyethylene in the benchmark, a particular focus was placed on thermal neutron scattering, both on the generation pipelines of ACE and HDF5 S(α,β) files and on the treatment of this type of scattering by the two different transport codes.

Speaker: Marta Campos Fornés (ATG Science & Engineering S.L.)

ICSBEP_concrete_labyrinth-Abstract.pdf

Wednesday 10 June

09:00 → 10:00 Neutron Source Facilities Design and Exploitation

Performance evaluation
Neutron/gamma diagnostics and calibration

09:00 Status and Recent Developments of IFMIF‑DONES: Neutronics Activities

The properties of materials for nuclear fusion applications are not yet fully understood under the neutron irradiation conditions foreseen in future fusion reactors. IFMIF DONES is designed to address this gap by providing neutron irradiation conditions representative of fusion environments. IFMIF DONES is an accelerator-based neutron source composed of a deuteron injector, a Radio Frequency Quadrupole, a Medium Energy Beam Transport line (MEBT), a Superconducting Radio Frequency accelerator, and a High Energy Beam Transport line (HEBT) to the Test Cell (TC). Together, these systems produce, accelerate, shape, and transport a deuteron beam to a liquid lithium target, generating fusion like neutrons for materials irradiation purposes.

Currently, the IFMIF DONES neutronics team is working in parallel on several key activities. These include: (a) rescoping and consolidating the knowledge transfer and results of the extended neutronics models available for the facility, based on the designs developed up to 2025 in the context of Eurofusion-WPENS collaboration; (b) providing detailed technical information and requirement definitions requested by in kind contributors (e.g., Croatia and Fusion for Energy); (c) supporting the transition of the MEBT, HEBT, and Test Cell—part of the Spanish contribution—into the construction phase; and (d) addressing the final design of the Main Building and site layout, whose construction phase is expected to begin shortly.

To manage these activities and accommodate new design developments, an intensive program of review and nuclear analysis is required, making use of the latest Eurofusion-WPENS results and targeting the licensing process with the Spanish regulatory authority (Consejo de Seguridad Nuclear - CSN). In parallel, the comprehensive characterization of the IFMIF DONES neutron source is a critical objective. To this end, nuclear data validation and diagnostic development are being pursued through the active promotion of complementary experimental rooms, considered as baseline components of the facility to enable such experimental activities in the near future.

This work presents the current status and recent advancements of the IFMIF DONES plant, with a focus on their impact on neutronics activities. In addition, ongoing efforts toward achieving a comprehensive characterization of the neutron field are discussed.

Speaker: Miguel Macias (IFMIF-DONES Consortium)

Abstract_2nd_Fusion_Neutronics_Meeting_MACIAS.pdf

09:20 Comprehensive Radiological Dose Mapping of the IFMIF-DONES Accelerator System During Operation and Shutdown

Radiological dose mapping in the IFMIF-DONES accelerator system poses a demanding neutronics problem because the radiation field is not governed by a single source term, but by the superposition of multiple contributions with different physical origins and spatial distributions. During beam-on, biological dose and absorbed dose in silicon are largely driven by neutrons and photons produced in deuteron interactions with structural materials along the accelerator line, together with source contributions associated with the lithium target system. During beam-off, the radiological field is determined by decay-photon sources arising from activation in irradiated components and surrounding structures. A consistent treatment of these source terms is therefore required to obtain reliable dose maps for shielding design, radiological accessibility, maintenance planning, and assessment of risks to sensitive equipment.
In this work, a comprehensive methodology is presented for the calculation of beam-on biological dose, beam-on absorbed dose in silicon, and shutdown residual biological dose distributions in the IFMIF-DONES accelerator system. The approach is based on detailed MCNP neutronics models combined with dedicated source treatments for the relevant prompt and decay radiation contributions. A major difficulty in accelerator radioprotection calculations involving light ions is that deuteron-induced nuclear reactions in matter produce a very low number of secondary particles per primary history, so that direct Monte Carlo transport of the resulting neutron and photon fields becomes statistically inefficient and often computationally prohibitive. This difficulty is especially relevant when secondary source production is distributed along extended beamline regions and must be coupled to complex facility geometries.
To address this problem, the UNED-developed source methodology srcUNED-Ac is used to generate and transport secondary-particle source terms associated with distributed deuteron losses along the accelerator line, enabling an efficient evaluation of neutron and photon fields arising from beam-loss interactions. Combined with the source contributions associated with the lithium target system and activation-induced decay sources, this methodology enables a consistent description of the radiological environment under both operational and shutdown conditions. The resulting spatial dose maps provide a comprehensive radiological characterization of the accelerator system and identify the most relevant regions for shielding optimization, ALARA-oriented access planning, maintenance strategy, and component protection. More broadly, the methodology demonstrates how advanced source-generation and transport treatments are required to perform statistically reliable dose mapping in fusion neutron source facilities.

Speaker: ANTONIO JESUS LOPEZ REVELLES (UNIVERSIDAD NACIONAL DE EDUCACION A DISTANCIA (UNED))

Abstract_ALZ_UNED.docx

09:40 High Intensity D-T Neutron Source HINEG and Its Applications

Neutron sources are the important experimental platform for the R&D of advanced nuclear energy systems, especially for the development for fusion systems. Series High Intensity D-T Steady Neutron Sources (HINEG) have been developed in China for different missions including neutronics design validation, materials & components irradiation test, nuclear waste burning and nuclear technology applications, etc. HINEG includes three phases: HINEG-I, HINEG-II and HINEG-III.
HINEG-II is a new D-T neutron source based on a high-voltage electrostatic accelerator, which is constructed in Chongqing and achieves an excellent neutron yield of 1.3×10¹³ n/s. Its outstanding performance is supported by three key core technologies: high-efficiency beam transport technology, which ensures stable extraction and transmission of high-current deuterium beams; a large, ultra-high-speed rotating target that can stably operate at 5000 rpm, combined with dynamic disturbance-enhanced heat dissipation technology to solve the problem of ultra-high heat flux density; and gradient nano crystallization epitaxial growth technology with a T/Ti ratio of 1.9, which improves tritium storage capacity and reaction efficiency. The device provides multi-type neutron spectra, including fusion-like ones, serving as a crucial experimental platform for fusion-related technical verifications, radiation damage mechanism research, and advanced reactor technology validation. Series experiments have been successfully completed on HINEG, including neutronics performance test of TBM blanket, measurement of leakage spectra from Pb and Pb-Bi, irradiation damage testing for laser crystal, etc.
Besides, the FDS Consortium has also developed a series of neutron sources for various uses, such as the Mini Neutron Generator (MINEG), Small Neutron Generator (SNEG), Compact Neutron Source (CONEG) and Volumetric Neutron Source (VNEG). Among them, SNEG is a D-D/D-T neutron source featuring with high neutron yield, compact design, and mobility. Its main body occupies an area of less than 1 m2, with a D-D source strength of 10¹⁰ n/s and a D-T source strength of 10¹² n/s. Due to its small size and portability, the SNEG neutron source is highly suitable for on-site calibration of detectors. VNEG is a Gas-dynamic Trap (GDT) based volumetric fusion neutron source with a neutron flux higher than 10¹⁵ n·cm⁻²·s⁻¹, it is an optimal platform for research on tritium breeding blankets of fusion reactors and material irradiation testing.
FDS provides an open platform that invites global research and applications in neutron source and fusion technology.

Speakers: Ms Qi Yang (Iternational Academy of Neutron Science), Mr Jun Gao (International Academy of Neutron Science), Mr Wei Wang (International Academy of Neutron Science)

Abstract_High Intensity D-T Neutron Source HINEG and Its Applications（Qi YANG 2026-03-30）.docx

10:00 → 10:40 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

10:00 Neutronics analyses of the VNS divertor and lower port

In recent years, a Volumetric Neutron Source (VNS) has been developed within EUROfusion with the aim of first demonstrating feasibility and then designing a machine that would reduce risk for the successful operation of teh demonstration fusion power plant DEMO. This would be achieved by designing and constructing a facility where crucial nuclear technologies, such as the tritium breeding blanket, can be tested and validated. The VNS is envisaged as a beam-driven, modestly sized tokamak which, due to its limited size, would be relatively quick to build, reasonably priced, and have tritium consumption low enough that it can be supplied by external sources, while still providing test conditions close to those expected in fusion power plants.

One of the challenges in integrating various systems is the lower port, where multiple functions must be consolidated into a design that meets all design requirements and limits. In this presentation, we describe our work on neutronics analyses in the area of the divertor and lower port. The focus was on reducing neutron-induced helium production in the divertor cooling pipes and investigating nuclear loads in the vacuum vessel below the divertor. The effect of introducing different shielding materials into the divertor and neighboring components was examined, and the impact of openings and gaps was quantified.

Speaker: Aljaž Čufar (Jožef Stefan Institute)

Abstract-2ndFNM_ACufar003.docx

10:20 Comprehensive Neutronics Analysis of Shielding Performance and Effects on Tokamak and building Structures in a Volumetric Neutron Source

For optimize nuclear fusion reactor designs, the need for detailed neutronics analysis is one of the important points, especially in terms of shielding performance and the overall impact on tokamak and facility safety. This study presents an extensive neutronics simulation of a Volumetric Neutron Source (VNS) focusing on the inboard area of a tokamak, coupled with a sector-based evaluation of the neutronics effects on the tokamak and its associated inside of building.
The inboard shielding effectiveness was thoroughly investigated through a layering method analysis using advanced shielding materials. Each layer of the shield blanket (SB) and vacuum vessel (VV) body was configured with shielding materials chosen based on their neutron flux, nuclear heating and dose rate properties. Various advanced shielding materials were applied to the design of shielding space to determine their impact on minimizing neutron flux and heating on ensuring that critical components, particularly the Toroidal Field Coils (TFC), are adequately protected against neutrons over the tokamak operational life.
Furthermore, the study extended beyond the tokamak to explore the neutronics impact on the inside of building. This comprehensive analysis performed 360° simulation of building model with a sector-based tokamak model to map neutronics results throughout the facility. This study could point out potential hazards in areas where additional shielding may be needed to protect equipment and personnel from the neutrons. The integration of these neutronics study findings into the fusion reactor design process is needed. By addressing direct and indirect neutronics impacts, this study supports the development of safer and more efficient design of devices. The outcomes can be enhanced understanding of complex neutronics interactions within tokamak but also contribute to the design and safety, realized for fusion development.

Speaker: Jin Hun Park (Karlsruhe Institute of Technology)

10:40 → 11:10 Coffee Break

11:10 → 12:30 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

11:10 Neutronics simulations for magnetic diagnostics in DEMO

As tokamak operation advances toward long-pulse and steady-state regimes, plasma-facing diagnostics will accumulate increasing nuclear loads over the lifetimes of the facilities. In such environments, the possibility of diagnostic failure due to neutron exposure cannot be excluded. Therefore, the design strategy for DEMO diagnostics has favoured robust front-end components compatible with remote handling. Among the set of diagnostics under study, magnetic sensors are the primary tool for plasma control, required to measure the local magnetic field distribution at numerous poloidal and toroidal locations. In-vessel pick-up coils and Hall sensors must be positioned between the breeding blanket and the vacuum vessel, where they will be exposed to a radiation environment with neutron fluences exceeding those expected in ITER by an order of magnitude or more.

This paper presents a comprehensive neutronics characterisation of the radiation environment at the anticipated locations of in-vessel and ex-vessel magnetic sensors in DEMO, for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) blanket configurations. MCNP simulations were performed using heterogeneous reference models of both blankets. Neutron and gamma fluxes, nuclear heating, dose rates, and displacements per atom (dpa) were calculated at 60 in-vessel and 60 ex-vessel sensor positions distributed poloidally around the plasma. The influence of the Magnetic Strip (MS) — a modular structure proposed to integrate the pick-up coils, Hall sensors and cable looms while ensuring compatibility with remote maintenance operations — was also assessed, through a parametric study in which the radial carving of the blanket required to accommodate the sensors was varied between 5 and 11 cm. The results show that carving the back of the blanket leads to a pronounced local increase in the nuclear loads at certain positions. Several structural and functional candidate materials were included in the simulations, enabling a complete characterisation of the radiation fields relevant to the design and integration of in-vessel magnetic diagnostics in DEMO.

Speaker: Raul Luís (Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa, Portugal)

RL_neutronics_magnetic_sensors_DEMO.docx

11:30 Preliminary design and assessment of a Fusion-Fission Hybrid TBM for Tokamak Devices

Fusion-fission hybrid reactors represent a promising pathway for future energy supply. However, the feasibility of this technological route has yet to be validated through a Test Blanket Module (TBM)-scale experiment, and comprehensive theoretical designs are still required to guide such experimental efforts. In this study, an innovative TBM design containing spent fuel from Pressurized Water Reactors was developed for integration into fusion devices. The design was comprehensively assessed using Monte Carlo methods. The analysis focused on key economic performance indicators, including the Tritium Breeding Ratio (TBR), multiplication of nuclear heat deposition, decay heat, and transmutation rates of major actinides, as well as critical safety parameters such as displacement per atom, the effective multiplication factor, nuclear heat removal, gas production rate, and fission product containment.
Results show that introducing spent fuel increases the overall neutron flux in the TBM. An optimized layout improves irradiation uniformity of the tritium breeder material, which in turn enhances its economic potential. At an appropriate volume fraction, spent fuel raises the TBR by nearly 2% relative to the design without spent fuel. The ratio of total nuclear heat deposition in the TBM to the total fusion energy entering the TBM is 458. Safety analyses confirm that the spent fuel remains deeply subcritical under all operating conditions, with a maximum keff below 0.2. The additional irradiation damage stays within acceptable limits for structural materials. Along with these performance improvements, the TBM design provides sufficient cooling space to safely remove nuclear heat generated during operation and decay heat after shutdown. Furthermore, at a fusion power of 10 MW, the fuel cladding effectively retains fission products and prevents their release. In summary, this study proposes a TBM design suitable for experimental validation in fusion–fission hybrid reactors, which is expected to accelerate the engineering application of this technology.

Speaker: xilong Tong (University of Science and Technology of China)

Abstract_Preliminary design and assessment of a Fusion-Fission Hybrid TBM for Tokamak Devices_Xilong.docx

11:50 Neutronics Activities at ENEA-Frascati: Experimental and Computational Developments

The neutronics activities carried out by ENEA in support of the ITER, DEMO, DTT, and EUROfusion programmes span a over a wide range of topics related to nuclear fusion technology. These activities include advanced three-dimensional neutronics analyses, activation calculations, and radiation shielding studies as well as experimental campaigs dedicated to the validation of codes and nuclear data, detectors and electronics design development.
The major activities in the ITER framework include nuclear studies supporting the integration of diagnostic equatorial ports (EP #2, #8 and #12), including the evaluation of nuclear loads on both the in-vessel and ex-vessel components, as well as comprehensive shutdown dose rate (SDDR) and activation assessments In addition, dedicated development of neutron source models and performance analyses have been carried out in support of the Radial Neutron Camera, with the aim of enabling two-dimensional neutron emissivity reconstruction.
Within EUROfusion, ENEA contributes to neutronics activities for DEMO and DEMO-LAR through analyses of the divertor and WCLL (water-cooled lithium lead) breeding blanket, with particular focus on tritium self-sufficiency and radiation shielding performance.
Significant effort is also devoted to the experimental validation of predictive tools and nuclear data, addressing key phenomena such as neutron streaming, shutdown dose rate (SDDR), and water activation. These studies, based on data from JET DT3 campaign, are complemented by benchmark activities on material activation measurements and dosimetry system development in DT radiation environments in support of ITER.
At FNG, ENEA leads and supports a wide range of benchmark shielding experiments on key materials (i.e. concrete, tungsten) as well as the development novel detectors and radiation hardness testing on electronics (RADNEXT framework, GENeuSIS project). Furthermore, a dedicated water loop has been designed and procured for the study and validation of ACP chemical behaviour and transport mechanism.
The new DTT neutronics model has been developed, integrating the latest updates in the machine design. Such model has been used to assess the impact of impurities on structural materials in terms of SDDR and it will be employed as a baseline configuration for the future nuclear analyses.

Speaker: Fabio Moro (ENEA, Nuclear Department)

FNM_abstract_Fabio.docx

12:10 Study of Activated Corrosion Products supporting the ITER and fusion pilot plant reactors: from neutronics to safety analyses and experiments

The wetted surfaces of water-cooled loops in nuclear reactors release corrosion products (CP) which, under neutron irradiation, become activated corrosion products (ACP), representing a major radiological concern. Their behavior is governed by complex multi-physics interactions involving neutronics, activation, corrosion, chemistry, and thermo-fluid dynamics. While ACP assessment in fission systems relies on well-established and extensively validated computational tools (e.g., OSCAR), fusion devices, such as ITER and fusion pilot plants, present new challenges due to different neutron spectra and operating scenarios, the presence of magnetic fields, the use of advanced materials (e.g., CuCrZr, Eurofer) and novel design of the cooling system. Therefore, dedicated analyses and validation are required to ensure accurate source-term evaluation, optimize occupational radiation exposure, and support waste management and maintenance strategies. Within the EUROfusion programme, several activities are currently ongoing to support the analyses, development and validation of computational tools for ACP assessment. A key objective is the validation of the OSCAR-Fusion code as well as other modelling approaches under ITER and pilot plants-relevant conditions. This work provides an overview of the ongoing activities within this research framework, focusing on the enhancement of the ACP calculation methodology mainly based on the MCNP, FISPACT-II and OSCAR-Fusion codes; the validation and modelling of ACP simulation tools and work-flow; and the support provided to dedicated experimental campaigns. These include the development of experimental water loops for fusion ACP studies (i.e., the ACP ENEA-FNG loop and the UKAEA corrosion loop), as well as corrosion experiments and measurements on ITER-grade CuCrZr samples carried out in the RINA testing loop. The main advances arising from these studies, as well as the key critical issues and the remaining open points affecting the uncertainty in ACP quantification, will be highlighted and critically discussed.

Speaker: Simone Noce (ENEA)

Abstract_FNM2026_Noce_v3.docx

12:30 → 13:10 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

12:30 Automated CAD to Neutronics Workflows with Stellarmesh

The development of an automated, parametric workflow to generate high-quality neutronics meshes from complex source geometries remains a key practical challenge in fusion neutronics. These workflows are particularly challenging given the need for element quality, metadata tracking, and mesh conformity throughout a multi-stage pipeline.

Stellarmesh is an open-source library developed at Thea Energy to address these challenges. It provides a unified API for the creation of MOAB, libMesh, and DAGMC meshes from source geometry in CadQuery, build123d, STEP, and BREP formats. In addition, Stellarmesh supports imprinting and merging of conformal geometry, native surface and volume meshing using Gmsh and OpenCASCADE, and the programmatic manipulation of .h5m files.

A key feature of Stellarmesh is its extensible backend architecture. While it ships by default with a fully open-source toolchain to mesh OpenCASCADE geometry, its unified backend API enables users to additionally integrate custom geometry and meshing backends. Given an API-compatible mesh file, it offers the same visualization, manipulation, and export functionality as the built-in backends.

This functionality has enabled Thea Energy to develop a separate backend to the Stellarmesh workflow that supports complex engineering models. These geometries, which are modeled by the engineering team in Siemens NX, are automatically exported to Parasolid files, meshed using the Simmetrix Simulation Modeling Suite, and imported into Stellarmesh using the API.

In this presentation, two workflows are showcased that demonstrate Thea Energy’s end-to-end process to create OpenMC models from CAD geometry. The first shows a sample blanket study using the open-source toolchain to illustrate how analysts can drive geometry generation, meshing, and neutron transport from a single Python file or environment. The second shows an analysis of a high-fidelity NX model, demonstrating how component metadata and geometric fidelity are preserved from CAD to OpenMC.

Stellarmesh is MIT licensed and available on both PyPI and conda-forge. This work shows that a well-designed and open-source meshing layer can substantially reduce the complexity of traditional CAD to neutronics pipelines and enable fully parametric, automated analysis.

Speaker: Alexander Koen (Thea Energy)

2026-04-02 2nd Fusion Neutronics Conference - Alexander Koen Abstract.docx

12:50 Development of parallel automatic void generation capabilities in GEOUNED conversion tool

Neutronics simulations in fusion facilities are done typically with Monte Carlo radiation transport tools like MCNP. While these tools use geometries based on Constructive Solid Geometry (CSG), fusion facilities are typically modelled as CAD models using the B-rep (boundary representation) system. The most successful approach to overcome this limitation is CAD model conversion into CSG models. In conversion tools, automatic void generation is one of the key features. Although engineering CAD tools are able to create the empty space that surrounds the CAD model, this approach is prohibitive for highly complex CAD models, like those of the ITER reactor, and therefore conversion tools have automatic void generation capabilities (García et al., 2021).

In this work it is presented, for GEOUNED conversion tool (Catalán et al., 2024), the parallelisation of the automatic void generation. Main design objectives were to achieve an optimal workload among several CPU cores and the progressive use of the parallel resources. Two parallel implementations were developed: block and progressive ones. Using a complex ITER CAD model, it was evaluated the GEOUNED computational parallel performance generating the void space. The two parallel implementations and a sequential one (without parallel capabilities) were evaluated. Finally, it was concluded that void space generation was faster with both parallel implementations compared with the sequential one, being the progressive parallel one the fastest. Also, the progressive parallel implementation achieved a better parallel workload distribution, a faster parallel data exchange and a progressive use of parallel resources.

Keywords: Conversion tools, Parallelisation, Automatic void generation, Particle Transport, Monte Carlo, CAD-based simulation

Acknowledgements: This work has been supported by UNED for the funding of the predoctoral contract (FPI). Also, this work has been made as part of the ETS Ingenieros Industriales-UNED doctoral program. The author also recognises Juan Pablo Catalán and Javier Sanz from UNED regarding the discussions about the development of the presented work.

Speaker: Juan García Bueno

Abstract_Garcia_Bueno_Juan.pdf

13:10 → 14:10 Lunch Time

14:10 → 15:10 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

14:10 A User‑Friendly OpenMC‑Based Workflow for CAD‑Driven Fusion Neutronics and Engineering Design

High‑reliability neutronics simulations are a critical component in the design and optimization of fusion systems. However, the complexity of modern Monte Carlo tools such as OpenMC often presents a significant barrier to their efficient use within fast, iterative engineering workflows. In this contribution, we present a user‑friendly application developed at Marvel Fusion that simplify and automates OpenMC‑based neutronics calculations, with the goal of enabling rapid and consistent engineering assessments by both neutronics specialists and non‑expert users.
The application provides an integrated workflow that enables rapid neutronics assessment directly of CAD‑based designs, with a particular emphasis on fusion‑relevant neutronic quantities. It allows our engineering team to efficiently evaluate the impact of geometry and material modifications on neutron multiplication, tritium breeding ratio (TBR), spatial neutron flux distributions, and neutron interaction rates in structural materials. In addition, the tool quantifies volumetric nuclear heating and the resulting power deposition, thereby supporting coupled neutronics thermal analyses. The application enables consistent, reproducible neutronics studies within fast engineering design iterations, making advanced Monte Carlo neutronics accessible to non‑expert users while preserving physical precision and reliability.
While the application has been developed to support laser‑driven fusion concepts at Marvel Fusion, its methodology is broadly applicable to fusion neutronics problems that require tight coupling between neutronic performance, thermal constraints, and CAD‑driven design changes. Overall, this work demonstrates how lowering the usability barrier of advanced Monte Carlo neutronics tools can significantly accelerate fusion engineering design cycles, enabling physics‑driven decisions to be made earlier, faster, and more reliably across multidisciplinary teams.

Speaker: Dr Jonathan Walg (marvel fusion)

Abstract KIT 070426.pdf

14:30 Implementation of a Deuterium-Lithium Neutron Source in OpenMC for IFMIF-DONES Neutron Calculation tasks

Speaker: Roman Afanasenko (Karlsruher Institut für Technologie (KIT))

14:50 Experimental verification and improvement of the TCV facility computational model

In this work, we describe how experiments where the energy response function and the angular response function of a neutron REM counter were modified, led to the improvement and validation of the computational model of the TCV facility. Due to an upgrade of the plasma heating system at TCV and a subsequent order-of-magnitude increase in neutron production, an upgrade of the radiological shield was needed. The upgraded radiological shield, made from polyethylene (PE), was designed using Monte Carlo neutron and gamma transport simulations and was installed in 2023.

During the design process of the radiological shield, the attenuation coefficient of the H*(10) neutron dose by PE was measured by placing the Lupin BF3 REM counter inside PE boxes of varying thickness. However, a significant discrepancy outside the statistical uncertainty was observed when these experiments were recreated with the MCNP code. This led to the realization that a significant amount of neutron moderators is missing in the MCNP model of the device, mainly the graphite protection tiles. After their addition, the relative ratio between calculations and experiments (C/E) for stacked PE boxes decreased from the range [1, 4.5] to the range [1, 2.5].

Additionally, the computational model of the TCV facility was validated using a shaded detector setup, where the Lupin BF3 counter was placed inside the PE box with one of the 6 sides left open, which enabled us to quantify the angular contribution to the neutron dose. A good agreement between experiments and calculations was observed with C/E ratios in the range [0.75, 2.0], for 10 measurements at 5 positions across the TCV facility. We propose a method with which these measurements can be used to determine the moisture content of the TCV hall concrete, which is expected to be a large source of error in dose calculations. Finally, we justify the use of the TCV experiments and the computational model we have developed as benchmark experiments by comparing them to existing benchmark experiments in fusion devices such as the JET neutron streaming experiments.

Speaker: Mark Fortuna (University of Ljubljana; "Jožef Stefan" Institute)

KIT_neutronics_meeting_abstract-Fortuna.pdf

15:10 → 15:40 Coffee Break

15:40 → 16:10 Invited Session

15:40 EUROfusion Stellarator Neutronics

Interest in stellarator-based fusion power plants has grown considerably within the fusion community, positioning stellarators as a promising alternative to tokamaks and motivating efforts toward their technological maturity and reactor readiness. Within this expanding stellarator ecosystem, the EUROfusion programme provides a coordinated, reactor-oriented, and physics-based framework for advancing stellarator design. In this context, stellarator neutronics has become a key discipline for evaluating radiation transport, shielding performance, nuclear heating, and material damage in complex three-dimensional reactor environments, where strong spatial non-uniformities and intricate magnetic and structural geometries play a dominant role.
Neutronic studies and design efforts must address the inherent complexity of stellarator configurations, including fully three-dimensional geometries, highly irregular spatial constraints such as periodic sector variations, concave–convex transitions, rotated ports, and tightly integrated coil systems. These aspects are critical not only for overall reactor performance assessment but also for supporting the development of subsystems such as breeding blanket (BB) technologies which depend critically on accurate neutron transport predictions.
Extensive work has focused on the Dual Coolant Lithium-Lead (DCLL) BB concept for the Helical-Axis Advanced Stellarator (HELIAS 5-B). Dedicated modelling frameworks such as HeliasGeom (UNED, Python) and SHANE (CIEMAT, Python + VBA CATIA) have enabled increasingly realistic reactor representations, evolving from simplified parametric models to variable radial build configurations to balance Tritium (T) breeding performance and coil shielding. Using these models, further optimization campaigns have been carried out to improve the T Breeding Ratio (TBR) while reducing radiation damage (dpa) under different FW design assumptions to facilitate remote maintenance. In addition, the integration of coil systems into simplified models has enabled detailed shielding analyses focused on radiation damage, neutron fluence, and nuclear heating, identifying vulnerable coil regions requiring enhanced protection against quenching.
More recently, a new generation of optimized quasi-isodynamic (QI) stellarator configurations, demonstrating reactor-relevant plasma performance have been computationally designed. Among the most advanced ones is the CIEMAT-QI4X configuration, for which new BB development activities have started, building on methodologies and experience from HELIAS 5-B studies. In parallel, a mesh-based workflow for rapidly converting CAD stellarator geometries into unstructured MCNP6 models (at KIT) is being successfully validated and benchmarked against CSG-based approaches using SHANE and HeliasGeom for CIEMAT-QI4X.

Speaker: Dr Iole Palermo (Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid, Spain)

Abstract_EUROfusion Stellarators Neutronics_FNM2026.pdf

16:10 → 17:40 Panel discussion on urgent fusion neutronics tasks for fusion energy realization. Directions for following which neutronics can accelerate the practical use of fusion energy.

Initial points for the panel:

1. The Tritium Breeding Challenge: Beyond the TBR > 1.0:
   The most urgent task for fusion energy realization is demonstrating Tritium Self-Sufficiency.

2. High-Fidelity "Digital Twins" for Radiation Damage:
   To accelerate the practical use of fusion, we must predict the lifetime of the First Wall and Vacuum Vessel without waiting decades for experimental data.

3. Neutronics of Advanced and Non-DT Fuels:
   While DT (Deuterium-Tritium) is the primary focus, the panel could explore how neutronics enables "cleaner" or more efficient cycles. Assess the neutronics of Cat-DD or Proton-Boron cycles.

4. Safety, Licensing, and Radwaste Classification:
   For a fusion plant to be "practical," it must be licensable by civilian nuclear authorities. Accurate assessment of Shutdown Dose Rate (SDDR) and Activation Analysis.

17:45 → 18:30 Bus transportation to Fusion Neutronics Dinner at "Brauhaus Kühler Krug"

Bus transportation from the FNM2026 Venue at KIT Campus North to the Fusion Neutronics Dinner at "Brauhaus Kühler Krug": https://brauhaus-karlsruhe.de/

18:30 → 21:30 Fusion Neutronics Dinner at "Brauhaus Kühler Krug" https://brauhaus-karlsruhe.de/

Thursday 11 June

09:00 → 10:00 Invited Session

09:00 A STEP on the path to commercial fusion: a nuclear analysis overview

The Spherical Tokamak for Energy Production (STEP) is an ambitious programme to generate net energy from fusion and to stimulate an industry that will help prove its commercial viability. It will achieve this by producing a prototype tokamak powerplant to provide energy to the grid.

The project is currently refining the overall plant architecture, plasma configuration and supporting technologies into a fully integrated design. Significant progress has been made on defining the machine geometry, validating the plasma operating scenarios, and developing the manufacturing and materials strategies required for a spherical tokamak operating at power reactor relevant conditions. Parallel workstreams are maturing the site and building consenting, licensing approach, digital engineering environment and supply chain capabilities.

Neutronics analysis plays a central, critical role in STEP’s design, informing component lifetime predictions, shielding optimisation, tritium breeding performance, nuclear heating, activation, and maintenance strategies. This presentation outlines the robust workflows and methodologies used within STEP to enable rapid design iteration, parametric exploration, and uncertainty quantification, while ensuring version control and that radiation environments are accurately characterised throughout the tokamak, blanket, shield, and auxiliary systems.

As the programme progresses and formal Development Consent Order (DCO) applications commence (required in the UK for nationally significant infrastructure projects), increasing attention is being directed beyond the tokamak itself to encompass supporting systems such as power and cooling infrastructure, as well as the overall building and site layout. This broader focus is necessary to enable an integrated assessment of all nuclear hazards.

In addition to addressing nuclear responses within the tokamak, this presentation will outline ongoing work to extend the nuclear analysis across the wider building complex, ensuring that radiation environment requirements and operational considerations are consistently understood and integrated throughout the entire facility.

This work has been funded by STEP Fusion, a major technology and infrastructure programme led by UK Fusion Energy Ltd (previously UKIFS), which aims to deliver the UK’s prototype fusion powerplant and a path to the commercial viability of fusion.

Speaker: Tim Eade (UKIFS)

STEP_Overview_abstract_v2.pdf

09:30 Gauss Fusion Neutronics Overview

Gauss Fusion GmbH has recently established an open-source software-based workflow for stellarator neutronics that enables comprehensive and reproducible simulations of full-device stellarator machines. The framework integrates established community tools, including STELLOPT for magnetic equilibrium optimization, ParaStell for homogenized geometry generation, Coreform Cubit© and Gmsh for models meshing, and OpenMC for neutron and photon transport simulations. This integrated approach provides a robust and user-friendly environment for neutronics analyses of complex three-dimensional stellarator geometries. A key challenge in stellarator design is the accurate treatment of non-axisymmetric machine configurations, particularly when assessing breeding blanket performance and shielding properties under realistic engineering constraints. The developed workflow streamlines geometry preparation, meshing, material assignment, and transport simulation, thereby improving reproducibility, and accelerating iterative design studies. As a next step, Gauss Fusion has recently started to work on analyses of engineering design of the breeding blanket modules paving the way to the high-fidelity neutronics output.
In this contribution, we present the achievements in the neutronics assessment of a Tritium Breeding Blanket (TBB) concept HEXA® for the future stellarator power plants. The analysis evaluates critical performance metrics such as tritium breeding ratio, tritium production maps, nuclear heating, and radiation damage indicators across the blanket, lifetime components, and superconducting magnets. We demonstrate the capability to support detailed performance evaluation and rapid parametric studies for HEXA®, providing input for thermal-mechanical and CFD analyses, tritium transport modelling (FESTIM), information on radiation damage and nuclear waste assessment.

Speaker: Dr Egor Vezhlev (gauss-fusion)

FNM_2026_Abstract_GFG_Neutronics_EV.pdf

10:00 → 10:40 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

10:00 First application of N1S code for calculation of dose due to movement of activated DONES components

The IFMIF-DONES accelerator will accelerate deuterons up to 40 MeV in order to generate fusion-like neutron fluxes for materials studies. Deuterons will be lost along the beamline, in scrapers and collimators, in the beam dump, and in d-Li reactions at the target. These deuterons and the neutrons produced in reactions may activate materials, particularly in components close to the beamline. These highly activated components, include the high-energy beam transport line (HEBT) scraper, high-flux test module (HFTM) and target assembly (TA). After irradiation in the DONES accelerator, these components will be removed, transported and stored, leading to accumulated dose in different areas of the building. Understanding of these doses is important for the purposes of safety and protection of equipment.
At UKAEA, the Novel 1-Step (N1S) code [1] has been developed for single-step shutdown dose rate (SDDR) calculations in MCNP. In this work, N1S has been modified to account for movement of cells after irradiation. This allows the calculation of accumulated dose around a transport path, accounting for the detailed geometry of the problem as well as parameters of the movement such as movement path and speeds. The N1S method is verified by cross-comparison of the results for DONES HEBT scraper removal scenario against previous calculations using the D1SUNED code [2]. The HFTM and TA scenarios present further detailed applications of the code.
[1] T. Eade et al., Fusion Eng. Des. 181, 113213 (2022)
[2] A. Lopez Revelles et al., EUROfusion IDM: EFDA_D_2RRC3L (2024)

This work has been carried out within the framework of the EUROfusion Consortium, funded by the European Union via the Euratom Research and Training Programme (Grant Agreement No 101052200 — EUROfusion). Views and opinions expressed are however those of the author(s) only and do not necessarily reflect those of the European Union or the European Commission. Neither the European Union nor the European Commission can be held responsible for them.
This work has been funded by the EPSRC Fusion Grant 2022/27 [grant number EP/W006839/1]. To obtain further information on the data and models underlying this paper please contact PublicationsManager@ukaea.uk. For the purpose of open access, the author(s) has applied a Creative Commons Attribution (CC BY) licence to any Author Accepted Manuscript version arising.

Speaker: Tom Berry (UKAEA)

Abstract_TBerry_DONES-Component-Movement_v1.1.pdf

10:20 Adapting the D1S Methodology for High-Fluence Fusion Environments with Tungsten: application to ITER

The decay of radioactive nuclides activated during the operation of nuclear fusion reactors represents one of the main safety concerns, leading to worker exposure during shutdown and maintenance phases as well as damage to critical electronics. For this, safety demonstrations of fusion reactors like ITER require a precise assessment of the Shut Down Dose Rate (SDDR). In recent years, the Direct-1-Step (D1S) methodology has been the most used due to its much greater computational speed compared to the Rigorous-2-Step (R2S) approach. However, when the assumptions underlying the D1S methodology no longer hold, e.g. when multi-step decay reactions occur in materials like tungsten, alternative approaches must be investigated. This work aims to explore and develop possible adaptations of the D1S methodology for the specific case of the ITER reactor, where tungsten (W) has been introduced as the First Wall (FW) material. The activation of tungsten triggers multi-step decays that make the standard D1S unusable for accurate In-Vessel SDDR calculations. The objective is to propose a modified methodology that maintains the advantages of D1S while ensuring the necessary accuracy for ITER's radiological safety, and to evaluate and understand the impact of FW material replacement on In-Vessel SDDR.

Speaker: Alberto Bittesnich (ATG Europe (F4E))

Abstract_template.docx

10:40 → 11:10 Coffee Break

11:10 → 12:30 Neutronics Tools, Nuclear Data and Workflow Integration

Geometry, radiation transport, activation, multiphysics tools.
Uncertainty quantification.
Variance reduction for MC.
Nuclear data development and experiments
Benchmarking and V&V
AI and ML use for neutronics

11:10 R2SUNED implicit stochastic uncertainty propagation scheme applied to JET

Nuclear analyses performed using Monte Carlo (MC) radiation transport codes are affected by stochastic uncertainty arising from the probabilistic sampling of particle histories. While formal mathematical convergence cannot be guaranteed, MC transport codes provide statistical estimators—such as the relative stochastic uncertainty—that allow the assessment of result reliability. Compliance with the tally statistical tests, including sufficiently low stochastic uncertainties, is therefore a mandatory requirement to ensure the numerical robustness of the simulation results.
Complex nuclear analyses may require the coupling of two Monte Carlo (MC) radiation transport simulations. The most widely adopted two–MC-step methodology is the Rigorous Two-Step (R2S) approach, commonly used for shutdown dose rate calculations. This methodology sequentially couples an MC neutron transport calculation, an activation calculation, and an MC photon transport calculation. Consequently, the statistical uncertainty introduced in the initial neutron transport stage propagates through the subsequent steps, affecting the photon transport results. However, the standard R2S methodology does not provide quantitative information on the impact of neutron stochastic uncertainty on the final response, thereby missing relevant information regarding the numerical reliability of the calculated results. This limitation is usually addressed by increasing the number of particles simulated during the neutron transport calculation in order to reduce its statistical uncertainty as much as possible, at the expense of a significantly increased computational cost.
Existing methodologies developed to address this issue require calculating the neutron flux covariance matrix to propagate the stochastic uncertainty in the neutron flux to the response. However, the size of this matrix makes its explicit calculation impractical for realistic applications. At the same time, simplified assumptions for its structure can lead to a significant overestimation of the resulting uncertainty, thereby increasing the overall computational cost.
In this work, we present the R2SUNED implicit stochastic uncertainty propagation scheme, which overcomes these limitations. The proposed scheme enables an efficient and accurate assessment of the contribution of neutron stochastic uncertainty to the final response, without relying on prohibitive covariance matrix calculations. The methodology has been applied to a JET shutdown dose rate analysis, demonstrating its applicability to real fusion-relevant problems and its contribution to improving the numerical reliability and robustness of R2S-based results.

Speaker: Javier Alguacil (UNED)

NeutronicMeeting Uncertainty Propagation.pdf

11:30 Recent Progress in Nuclear Response Cross-Section Research at Harbin Engineering University

Nuclear response cross-sections, including KERMA (Kinetic Energy Release in Materials) factor, displacement damage cross-section and gas production cross-section, are critical nuclear data for nuclear heating, atomic displacement and H/He gas production. KERMA factor and displacement damage cross-section are researched at Harbin Engineering University recently for some potential problems.
Concerning neutron KERMA factors, values calculated using nuclear data processing codes often show erroneous results and lack physical consistency. This is because the energy balance in evaluated nuclear data libraries cannot be ensured, or the processing methods employed in the codes are inappropriate. To provide key information for the improvement and correction of evaluated nuclear data libraries, neutron KERMA factors for major reaction types—including scattering, fission, absorption, and radiative capture reactions—are calculated and analyzed based on the FENDL-3.2 and JENDL-5 libraries. All unphysical negative values in these reactions are identified, classified, and collated. The causes of the erroneous results are analyzed, and suggestions are proposed for revising the data in the evaluated nuclear data libraries.
For displacement damage cross-section, displacement function model is the key to its accuracy calculation. Traditional displacement function model cannot describes the whole and real physical processes, leading to precision loss. To obtain displacement damage cross-section correctly, an improved Athermal Recombination Corrected (i-ARC) DPA model is proposed, which uses the power function model near the average atomic displacement threshold energy and original ARC-DPA model in the other energy part. The iARC-DPA model has been used in the calculation of displacement damage cross sections for different incident particles, including electron, neutron and charged particles. The displacement damage cross sections of different particles are calculated and compared with the experimental data. Numerical results show that the iARC-DPA model largely improves the accuracy of displacement damage cross section calculation.
In conclusion, some researches on KERMA factors and displacement damage cross-sections are performed in the recent years at Harbin Engineering University to provide more accuracy nuclear response cross-sections. Nevertheless, much work remains to be done in the near future. For example, the specific causes leading to erroneous KERMA factors need to be identified, and more physically consistent KERMA factors should be provided. In addition, the differential cross sections for different incident particles still need to be re-evaluated to obtain more accurate displacement damage cross sections.

Speaker: Prof. Wen Yin (Harbin Engneering University)

Recent Progress in Nuclear Response Cross-Section Research at Harbin Engineering University.pdf

11:50 Building programmable study platform for Activated Corrosion Products assessment

Activation Corrosion Products (ACP) are one of the more challenging aspects of radiation protection assessment for the safety analysis nuclear installations. The generation, transport, and deposition of ACP in cooling circuits require powerful simulation solvers capable of representing the interplay of different physical phenomena raging from thermal-hydraulic, chemistry, neutron activation, and radioactive decay. At the same time, the complexity of the design of cooling circuits, the harsh and peculiar radiation environments, and the peculiarity of the materials and operational scenarios of nuclear fusion machines demands the possibility to perform parametric studies, define alternative irradiation modelling, and to combine the results of the ACP simulation with subsequent post-processing and additional side calculations.
While in the framework of fusion nuclear research several initiatives and ACP calculation strategies have been proposed, the OSCAR-Fusion code, developed by the CEA with the support of EDF and Framatome, is currently considered the reference, especially for the studies in support of the ITER safety demonstration.
The goal of this presentation is to provide an overview of software architecture strategies to construct a programmable study platform for Activated Corrosion Products assessment based on the experience acquired by ENEA in the framework of EUROfusion studies in support of DEMO and ITER. By starting with the point of view of the engineer in charge of the design and safety assessment and by considering the typical numerical modeling strategy for the ACP simulation, the components in charge of defining the study platform will be outlined.
The platform under development is based on the separation of responsibility following the three-iter architecture. The presentation layer exposes the interface on which the engineer can define the ACP study by representing the circuit in the form of control volumes with materials and geometries associated, and the typology of the study to be performed, and the respective tools to perform meaningful post-processing. The data layer is composed of a set of plugins capable of serializing the internal data structures into the actual input decks for the target calculation codes, e.g. OSCAR-Fusion for ACP simulation, FISPACT or OpenMC for inventory calculations, and data science structures capable of representing and processing the output inventories to produce meaningful aggregated results. The application layer implements the business logic capable of transforming the engineering needs to the actual simulations to be done and vice versa. Special emphasis has been placed on the development of the platform as a programmable set of high-level Python classes, so that the final user may be able to define a customized workflow.
An example of application to the studies previously done at ENEA, e.g. the ITER WCLL and/or the Frascati ACP loop, will be provided.

Speaker: Dr Alberto Previti (ENEA)

Abstract_AlbertoPreviti_ACPplatform.pdf

12:10 Time-Aware Conformal Uncertainty Quantification with Physics-Consistent Projection for Radioactivity-Related Time-Series Prediction in Fusion Blankets

Machine-learning surrogates for coupled neutronics–activation calculations (OpenMC + FISPACT-II) reduce per-sample evaluation time from ~40 min to below one second, making rapid blanket design iteration feasible. Point predictions, however, do not suffice for safety-critical decisions. Activation inventory, decay heat, and contact dose rate span more than six orders of magnitude across time scales from hours to $10^5$ years; credible uncertainty intervals are therefore indispensable.

Standard conformal prediction (CP) produces miscalibrated intervals in this setting because (i) predictive residuals are non-stationary across irradiation and cooling regimes, and (ii) physical constraints—nonnegativity and monotonic decay—invalidate coverage guarantees when enforced as post-hoc corrections. We present SC-PIML+, a physics-constrained conformal framework comprising: (1) physics-informed feature engineering with capped half-life detrending; (2) time- and phase-stratified nonconformity scoring; (3) a convex projection operator preserving nonnegativity and smoothness in log-space, applied before calibration scoring to maintain coverage validity; and (4) a cross-conformal (CV+) extension with half-life-adaptive projection policy that increases effective calibration sample size by ~30×.

We evaluate SC-PIML+ on 224 CFETR PbLi blanket geometries with 15 prediction targets in four categories: breeding isotopes ($^3$H, $^6$Li, $^7$Li), aggregate safety indicators (decay heat, total activity, contact dose rate), PbLi-origin activation products ($^{210}$Po, $^{210}$Bi, $^{203}$Hg, $^{204}$Tl, $^{207}$Bi, $^{203}$Pb), and steel-origin activation products ($^{54}$Mn, $^{60}$Co, $^{55}$Fe). In a six-method benchmark, SC-PIML+ is the only method that meets nominal 95% prediction interval coverage on all 15 targets; Split Conformal covers 12, Conformalized Quantile Regression 11, and the base SC-PIML 13. Across 20 independent random data partitions the method passes 10.5 ± 2.4 targets on average; the remaining spread stems from finite-sample calibration ($n_\text{cal} \approx 30$), not from method instability. Because the conformal calibration is decoupled from the base learner, gradient-boosted trees, MLPs, and LSTMs all run through the same pipeline without modification. An operational out-of-distribution detection metric flags extrapolation cases for automatic fallback to direct FISPACT-II evaluation.

Speaker: Xiaokang Zhang (Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences)

12:30 → 13:10 Neutron Source Facilities Design and Exploitation

Performance evaluation
Neutron/gamma diagnostics and calibration

12:30 Estimation of the STUMM-DONES detector's responses using the MC method

The Start-Up Monitoring Module (STUMM) is currently under development as a dedicated calibration and diagnostic system intended for the commissioning stage of IFMIF-DONES. The module will be positioned in proximity to the neutron source, within the irradiation zone used for material testing. Its main purpose is to enable accurate characterization of neutron source parameters and prevailing irradiation conditions.
The STUMM system is designed to incorporate around 240 detectors organized within specialized rigs. Its diagnostic configuration comprises eight Rabbit Systems (RS) for measuring thermal, epithermal, and fast neutron fluxes. Additionally, it includes 66 pairs of Micro-Fission Chambers equipped with U-238 and ionization chambers (amounting to 132 detectors) dedicated to fast neutron flux monitoring. For thermal and epithermal neutron measurements, 36 Micro-Fission Chambers with U-235 or alternatively Self-Powered Neutron Detectors (SPND) are planned. 28 gamma thermometers further complement the system for evaluating nuclear heating, and 44 thermocouples for temperature measurements.
Due to the extreme operating environment—high neutron fluxes and mixed radiation fields—the detectors must ensure strong radiation resistance while avoiding signal saturation. Although largely based on commercial technologies, several components require adaptation to meet IFMIF-DONES conditions.
This work presents studies of the STUMM system response under representative irradiation scenarios. Both active detectors (MFCs, ionization chambers, gamma thermometers) and the Rabbit System, treated as a semi-passive technique, are analyzed. The selection of activation materials and optimization of detector configurations are discussed. The results support validation of the diagnostic concept and reliable neutron source characterization during commissioning.

Speaker: Urszula Wiącek (Institute of Nuclear Physics Polish Academy of Sciences, Radzikowskiego 152, PL-31342 Krakow, Poland)

UWiacek_Abstract.docx

12:50 Assessment of irradiation modules diagnostics and capsule filling in IFMIF-DONES

The materials proposed for nuclear fusion applications have yet to be fully characterized under the neutron irradiation conditions expected in future fusion reactors. IFMIF DONES (International Fusion Materials Irradiation Facility – DEMO Oriented NEutron Source) has been conceived to address this challenge by irradiating materials under neutron conditions representative of those found in fusion environments. In IFMIF DONES, neutrons will be produced through the interaction of a 40 MeV, 125 mA deuteron beam, with a thick liquid lithium target, where deuteron–lithium reactions generate a high intensity neutron flux.

Downstream of the lithium target, different modules will be placed for the commissioning and operation phase. These modules will host diagnostics to machine protection, beam monitoring and beam characterization. Among the diagnostics currently considered are ionization chambers, fission chambers, and self powered neutron detectors. To better understand the behaviour of these detectors in the IFMIF DONES environment, theoretical sensitivity studies have been carried out. These studies aim to gain deeper insight into the influence of the detector design on the estimation of the induced current.

Moreover, to mitigate temperature gradients induced by gaps between specimens, the irradiation module capsules and diagnostics will be filled with a conductive material to ensure thermal homogenization. At present, sodium has been selected for this purpose, as it behaves as a liquid metal at irradiation temperature. However, sodium is corrosive to the specimen materials, which motivates the need to investigate, from the neutronic point of view, alternative filling materials in solid, liquid and gas states that can provide adequate thermal performance while ensuring material compatibility and without jeopardizing the neutron doses received by the specimens.

The assessment of the different designs and scenarios was carried out using calculations obtained from different Monte Carlo simulation codes.

Speaker: Irene Álvarez Castro

Abstractv2_IAZ.docx

13:10 → 14:10 Lunch Time

Canteen's Menu online: https://cga.se4.kit.edu/downloads/CGA/Speiseplan_englisch.pdf

14:10 → 15:10 Neutron Source Facilities Design and Exploitation

Performance evaluation
Neutron/gamma diagnostics and calibration

14:10 Neutron irradiation conditions in the IFMIF-DONES Test Cell and Complementary Experiments Room R160

This work presents computational neutronic analyses performed for the safety of the IFMIF-DONES Target Assembly components and parametric analyses of the Neutron Beam Tube and Shutter (NBT&S) diameter to increase the collimated neutron flux at the entrance to Complementary Experiments Room 160 (CER R160). The IFMIF-DONES project is under construction in Spain. The radiation transport parallel computations have been performed on the CPU-based partition of the Pitagora supercomputer hosted by CINECA in the framework of the MCHIFI (Monte Carlo High Fidelity) HPC EUROfusion project. Neutron spectra and atomic and nuclear responses have been calculated using the Monte Carlo MCNP code. The D+ ions are accelerated in IFMIF-DONES up to 40 MeV energy and 125 mA current. Heating calculations [1] in liquid Li and steel at the area of the deuteron (D+) footprint require the inclusion of the heat contributions of charged D+ ion particles. The integral heating calculations in IFMIF-DONES Test Cell (TC) components reveal that D+ energy deposition in liquid Li at the thin Bragg peak, with a D+ beam footprint area of 20x5 cm2, contributes 97% of the total heating in the whole Test Cell volume, delivering 5 MW heat power of D+ beam to liquid lithium.

A study of neutron flux growth at the entrance to CER due to increasing the NBT&S diameter has been performed. The neutron flux isolines are plotted and compared for three variants of the NBT&S diameter: 15 cm, 21 cm, and 30 cm. The illustration of how the streaming along the NBT&S depends on the aperture diameter is presented. Given the current design of NBT&S, the diameter could be increased from 15 cm to approximately 21 cm, making the total neutron flux grow to 13% at the CER entrance. The principal flux growth is defined by the characteristic dimension of the neutron source size projection of ~45 cm at the entrance from TC to the NBT&S.

References
[1] A. Serikov, et al., “Computational neutronics analyses of deuteron interactions with lithium target in IFMIF-DONES for fusion applications,” 5th Fusion HPC Workshop, online, November 21-22, 2024, https://doi.org/10.13140/RG.2.2.32675.57128, https://www.youtube.com/watch?v=uXF0BpS_eyk&t=6179s

Speaker: Dr Arkady Serikov (Karlsruhe Institute of Technology (KIT))

Abs_Serikov_v1.docx

14:30 Radiation Transport Analysis of Beam Loss in the LIPAc Superconducting RF Accelerator

As part of the IFMIF/EVEDA project, the Linear IFMIF Prototype Accelerator (LIPAc) was constructed at Rokkasho, Japan, and a Superconducting Radio-Frequency (SRF) accelerator has been installed for the upcoming 9 MeV deuteron beam campaign. In the previous 5 MeV beam experiments, radiation transport analyses using MCNP relied on manually created geometry models including only major components. However, beam losses in the SRF are critical due to its very small acceptance (10 W total), and accurate evaluation of neutron and photon transport is required for reliable interpretation of signals from beam loss monitors installed around the SRF.
Because of the complex structure of the SRF, manual MCNP modeling has significant limitations. In this study, the effectiveness of automatically converting SRF CAD data into a detailed MCNP model was investigated using the open-source conversion tool GEOUNED. A preliminary neutron transport analysis was performed using the converted MCNP model, assuming a beam loss at the center of the SRF beamline represented by a 1 MeV neutron point source. The detailed model reproduces internal SRF components, including solenoids and cavities, resulting in complex neutron interactions and a non-uniform flux distribution. These results demonstrate that accurate identification of beam loss locations and intensities requires radiation transport analyses that explicitly account for the internal SRF structures.

Speaker: Kohki Kumagai (QST)

(abstract) Radiation Transport Analysis of Beam Loss in the LIPAc Superconducting RF Accelerator_260408_R1.pdf

14:50 Investigation of proton-induced 7Li nuclear data for fusion neutron facilities

Lithium targets are commonly employed to produce “fusion-relevant” neutrons in fusion technology validation facilities, where proton-lithium (p-Li) reactions serve as the primary neutron source. Accurate neutron production cross sections are critical for facility designs, shielding calculations, and dosimetry. However, the quality and consistency of available proton-induced lithium nuclear data remain insufficiently assessed compared to neutron-induced ones.
This study evaluated the quality of neutron production data in p-7Li reactions with the p-Li experiment at the quasi monoenergetic neutron generator at the Nuclear Physics Institute of the Czech Academy of Sciences, Czech Republic [1]. The MCNP6.2 calculation results with the official ACE files of JENDL-5 and ENDF/B-VIII.0 were significantly different from the measured results.
1) The neutron spectra calculated with JENDL-5 had similar shapes to the measured ones, but the absolute values were different.
2) The neutron spectra calculated with ENDF/B-VIII.0 had different shapes and absolute values from the measured ones.
The neutron production cross section of 7Li in ENDF/B-VIII.0 is much smaller above 10 MeV than those in JENDL-5 because of lack of the (p,n+He3+He4) reaction, etc., which is considered to be the reason of the difference between the calculated neutron spectra with JENDL-5 and ENDF/B-VIII.0.
We have also studied the p-Li experiment at the Cyclotron and Radioisotope Center of Tohoku University, Japan [2]. The results will be presented in the workshop.

References
[1] Zdeněk Matěj, et al., The methodology for validation of cross sections in quasi mono energetic neutron field, Nuclear Instruments and Methods in Physics Research A 1040 (2022) 167075.
[2] Yoshitomo Uwamino, et al., High-energy p-Li neutron field for activation experiment, Nuclear Instruments and Methods in Physics Research A 389 (1997) 463-473.

Speaker: Saerom Kwon (National Institutes for Quantum Science and Technology (QST))

abstract_template_Neutronics2026_pLi_SKN3.docx

15:10 → 15:40 Coffee Break

15:40 → 16:40 Neutron Source Facilities Design and Exploitation

Performance evaluation
Neutron/gamma diagnostics and calibration

15:40 Assessment of the Neutron Generator emission parameters for the in situ calibration of ITER neutron diagnostics

The main goal of ITER operation is demonstration of fusion power reaching values up to 10 times greater than the supplied plasma heating power. This demonstration relies on the set of neutron diagnostics capable of measuring the neutron yield with 10% accuracy and of providing reliable data stream for regulatory purposes. The specified accuracy is planned to be achieved through in situ neutron calibration with a powerful sealed tube D-T neutron generator (NG) with the yield of up to 1e11 n/s as a source. For operation without tritium, a similar calibration is provisioned with a D-D NG (yield ~1e9 n/s). Source definition for neutron transport analysis of these campaigns is crucial and can only be done in conjunction with model validation in experiments with fusion neutrons.

We demonstrate the results of neutron flux and spectrum anisotropy measurements using sets of diamond detectors, fast scintillators, fission chambers, boron counters and activation samples. Some of the proposed detectors are shown to be suitable as a part of the NG monitoring system. The use of high sensitivity neutron counters allows to characterize the source from directions in the shadow of the irradiation unit. The model of the neutron source demonstrates good accuracy in the frontal sphere of the NG, while the discrepancies arising from NG inner contents constitute a systematic impact on the neutron flux and spectrum in the back part of the NG. We quantitatively discuss the shadow cone approach to better characterize the scattered neutron flux component during the measurements.

Current work is supported under the Task Agreement IO/25/TA/4500000251 as of 12.12.2025 between ITER Organization and Institution “Project Center “ITER”. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

Speaker: Mr Aleksei Pankratenko (ITER RF DA)

fnm-26-abstract-v2.docx

16:00 The TU-Dresden Neutron Generator

The TU Dresden Neutron Generator (TUD-NG) is an accelerator-driven deuterium–tritium (DT) neutron source designed for fusion-relevant irradiation experiments. The facility delivers deuterium beams of several milliampere onto a water-cooled tritium target, producing neutrons via the 3H(d,n)4He reaction with energies around 14 MeV. The source is designed to reach neutron yields of up to 1012 n/s with nearly isotropic emission, making it the most intense DT neutron sources in Europe.

Originally commissioned in 2003, the facility has been used intermittently for applications in fusion research, astroparticle physics, and nuclear astrophysics. Current efforts aim to significantly expand its operation and transform the TUD-NG into a dedicated user facility for German and European fusion research.

The experimental infrastructure enables irradiation studies of materials and components under fusion-relevant neutron spectra, activation measurements for the validation of nuclear data and transport simulations, and investigations of key processes such as tritium breeding. By providing flexible and comparatively rapid access to 14 MeV neutrons, the TUD-NG is positioned to bridge the gap between small laboratory sources and future large-scale facilities.

In this contribution, the design and performance characteristics of TUD-NG are presented, together with plans and ongoing developments toward user operations.

Speaker: Dr Bjoern Lehnert (TU-Dresden)

Abstract_blehnert.docx

16:20 Commissioning and hydrogen test of the ion source for Sorgentina-RF

In late 80s-early 90s, two projects of accelerator-based 14-MeV neutron sources were conceived at the ENEA Research Center of Frascati, i.e., the Frascati Neutron Generator (FNG) and SORGENTINA, in response to specific needs of the scientific community involved in Controlled Thermonuclear Fusion. FNG (neutron strength up to 1.5x1011 s-1) was mainly to support the development and benchmark of neutronics codes, methods and cross-sections relevant for the nuclear analysis of fusion machines, while SORGENTINA (neutron strength up to 7x1015 s-1) was to support studies on damage and property variation of fusion materials under higher neutron fluence.
FNG was then built and has operated for over thirty years in support of primary fusion neutronics experiments, while the SORGENTINA project has been reconsidered only in recent years at a smaller scale, i.e., 1014 s-1 (a.k.a. SORGENTINA-RF). SORGENTINA-RF is a 250 kW neutron source at his first stage of development and mainly funded to demonstrate the production of radio-tracers for medical diagnostics. This stage foresees the construction of the ion source and accelerator column and the rotating target where fusion reactions occur. In the ion source, a mixed positive ion beam of D+ and T+ (833 mA in total at the target, half of each) is produced in the RF plasma chamber of the ion source equipped with an antenna (in combination with a DC solenoidal magnetic to improve significantly the ion production), then extracted and accelerated up to 50 keV by a 4-electrode multi-aperture extraction system and then to 300 keV in the accelerating column. The present work aims at illustrating the machine performance and the recent commissioning and tests in hydrogen with a proper ion beam dump at the ENEA Research Center of Brasimone. Criticalities and next steps toward the nuclear phase of the machine are highlighted as well.

Speaker: Dr Nicola Fonnesu (ENEA)

Abstract FNM_2026_fonnesu.docx

Friday 12 June

09:00 → 10:40 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

09:00 Activation Analysis and Radiological Assessment of the ITER X-Ray Crystal Survey Spectrometer for Interspace Support Structure

The monitoring of plasma impurity influx in ITER is facilitated by the X-Ray Crystal Spectrometer-Survey (XRCS-Survey) located in Equatorial Port-11. To ensure high-sensitivity X-ray collection, this diagnostic utilizes a windowless line-of-sight; however, this open optical path creates a direct path for significant neutron streaming toward the spectrometer assembly. This study evaluates the resulting neutron-induced material activation to establish the technical foundation for radiation zoning, shielding optimization, and personnel safety protocols. As ITER transitions from low-power commissioning toward the SA-2 phase, the diagnostic environment shifts from non-nuclear validation to the high-intensity performance requirements of burning plasma under full Deuterium-Tritium (DT) conditions. To address these challenges, a comprehensive material activation study was conducted using a dual-stage computational workflow, coupling high-fidelity neutron transport results with the FISPACT activation code. Utilizing vitamin-j 175-group neutron flux files, scenario-specific irradiation and cooling schedules, the analysis identifies critical nuclide inventories, total activity, decay heat, and dose rates essential for radwaste classification and maintenance access control. Further analysis on the new irradiation scenario is ongoing for XRCS-Survey components, covering both the Interspace Support Structure (ISS) and Port Cell Support Structure (PCSS). Collectively, these results provide the validated data required for optimizing design, defining radiation zones, and ensuring the long-term safety of operations and decommissioning for the XRCS-Survey system.

Speaker: Bhargav Soni (ITER-India, Institute for Plasma Research)

Bhargav_soni_FNM_2026.docx

09:20 Comprehensive modeling of activated water radiation sources of ITER Tokamak Cooling Water System

The activated water has been identified as a relevant radiation source in fusion facilities during plasma operation. In ITER, cooling water is activated near first-wall components due to the intense neutron exposure and carries the produced radioisotopes far from the irradiation region. This phenomenon generates complex radiation fields outside the bioshield, directly impacting on the design of radiation shielding, radiological zoning, and licensing of the facility in regions where plasma neutrons are negligible. Hence, the accurate characterization of the activated water radiation source and its impact on the entire facility is critical for the project.
The Tokamak Cooling Water System (TCWS) is itself an enormous and highly complex network comprising multiple circuits organized in closed loops with varying extension and neutron exposure levels depending on the component of the tokamak that they cool. This results in a highly heterogeneous thermohydraulic network where radionuclides concentrations vary in several orders of magnitude. The characterization of the radionuclide production, decay and transport, and the associated dose fields is, therefore, a problem coupling neutronics and thermohydraulics.
In previous assessments of the TCWS activated water radiation source, pure water was assumed as coolant, resulting in the production of dominant but short-lived nuclides such as 16N and 17N. However, the implementation of realistic water chemistry to control the pH and minimize corrosion products introduces new activation pathways leading to the production of longer-lived radionuclides. The decay of these radioisotopes becomes relevant after a few minutes of transient time. Consequently, several TCWS sub-circuits that were previously disregarded because of the rapid decay of short-lived nuclides must now be reassessed.
To address these challenges, the TCWS has been modeled from scratch according to the latest designs. All principal and auxiliary cooling loops containing water during plasma operation have been considered. Updated thermo-hydraulic parameters to accurately describe flow rates and residence times have been used to compute the radionuclide activities in each pipe segment. The resulting geometry, composed of approximately 35,000 individual solids including pipes, water cells, filters and tanks, has been fully integrated into the ITER MCNP Full Model describing the entire facility. This integration enables a realistic evaluation of radiation transport from activated water sources within the global ITER radiation environment. A comprehensive nuclide inventory has been considered. Gamma and neutron emitting radiation sources have been generated for each circuit and isotope independently using the FLUNED-SL toolkit, and subsequently analyzed with D1SUNED to obtain the corresponding dose distributions.
The outcome of this work is a comprehensive, fully circuit-resolved, time-dependent and chemistry-consistent radiation source of ITER activated water during plasma operation. The results provide important input for the new release of ITER radiation maps supporting the forthcoming update of the ITER Safety Report.

Speaker: Dr Pablo Martínez Albertos (UNED)

FNM26_abstract_PMZ.pdf

09:40 From Tokamak to Research Reactor: Replicating JET Water Activation Experiment at KATANA Facility

Water activation plays a key role in fusion facilities, contributing to radiation fields, shutdown dose rates, and overall operational safety. High-energy neutrons produced during fusion reactions interact with oxygen and other elements in the coolant, generating short-lived activation products such as N-16 and N-17. These isotopes emit high-energy gamma radiation and neutrons, affecting radiation levels during operation and shortly after shutdown, and thus influencing shielding design, maintenance planning, and safety assessments.

The JET DTE3 campaign has provided valuable experimental data on neutron-induced activation in cooling water under fusion conditions. Water activation in the JET cooling circuit was measured during more than 1,500 JET pulses. Reproducing such complex experiments in smaller, controlled environments is essential for validating activation calculations and supporting detector development. In this work, three JET water activation scenarios were replicated at the KATANA water activation facility using pulses from the JSI TRIGA Mark II reactor.

The approach focuses on replicating the time-dependent neutron irradiation conditions observed in JET by using short, high-power TRIGA pulses. The KATANA facility enables controlled irradiation of water and its transport to the measurement volume. By adjusting irradiation parameters, the study aims to reproduce the water activation observed in the JET tokamak’s cooling circuit. Computational support with the KATANA activation tool was used to model the irradiation scenarios and predict activation responses for comparison with experimental results.

Although differences in neutron energy spectra between fission and fusion sources are known, the cross section for the O-17(n,p)N-17 reaction, with its threshold (Eₜᵣₑ ≈ 9 MeV), excludes the part of the spectrum where the two differ most. This allows the activation behaviour to be effectively approximated. Experimental results are compared with JET measurements and activation simulations to evaluate the extent to which the activation behaviour can be reproduced.

This work highlights the potential of research reactor facilities as flexible and cost-effective platforms for fusion neutronics studies, enabling experimental validation and supporting future fusion reactor development.

Speaker: Julijan Peric (Jozef Stefan Institute - JSI)

26_04_07_J_Peric_FNM_V2.docx

10:00 Evaluation of SRO mode-0 conditions and TSSI design

The Start of Research Operation (SRO) is the first nuclear operational phase of ITER, as defined in the ITER baseline research plan. During this phase, hydrogen and deuterium plasmas are operated at full magnetic field and plasma current, but with limited neutron production, with the primary objective of demonstrating machine integrity and nuclear safety prior to deuterium–tritium operation. The machine configuration and associated systems during SRO differ significantly from those of subsequent operational phases.
Fusion for Energy (F4E) has performed preliminary engineering assessments to evaluate radiation conditions during SRO. The objective of these studies is to establish shielding design requirements for the Structures, Systems and Components (SSCs) present during operation, with particular focus on the Temporary SRO Shielding Items (TSSI). These shielding provisions must ensure compliance with occupational dose limits and support safe assembly operations.
The production of these neutronics assessments is technically challenging due to the complexity of the computational models, the deep-penetration nature of the problem, and the extent of modifications required with respect to the reference models. These assessments were performed using the novel Gitronics methodology to track the changes required and the different configurations.
Neutron fields were initially calculated inside the bioshield using an ad hoc modified version of the E‑Lite model. In addition to implementing the SRO-specific configuration changes, the model was reduced to sectors 1, 2, and 3 to improve computational efficiency. The in-bioshield simulation provided neutron fluxes and energy spectra at the TSSI locations in the ports and top lid. Responses were subsequently used as source terms in simplified one-dimensional models to perform a scoping study aimed at identifying optimal port TSSI thicknesses and material compositions to feed the mechanical design. In parallel, a surface source (RSSA) was generated at the bioshield boundary and employed as input to a second simulation based on a modified Tokamak Complex model using the SRCUNED code. Also in this case, the Neutral Beam (NB) cell was adapted to reflect the SRO configuration, and neutron fluxes and spectra were recorded at the North Wall TSSI locations to conduct an additional scoping study for those items.
Once the optimal TSSI designs were identified, they were implemented in both the in-bioshield and Tokamak Complex models. Then assessments were reperformed with both models to confirm the expected radiation conditions for the SRO configuration. To conclude, final comprehensive full sectors simulations are planned along the summer.

Speaker: Alvaro Cubi (F4E)

Abstract_SRO_mode_0.pdf

10:20 Neutron Transport Analysis for ITER DNFM in situ Calibration in the Tokamak Environment

ITER Divertor Neutron Flux Monitor (DNFM) is designed to measure the total neutron yield and the fusion power in a wide dynamic range (1014 to 3x1020 n/s). Each DNFM detector unit is comprised of 235U and 238U fission chambers with independent electrode systems coated with either 500, 50 or 5 mg of uranium, allowing it to cover the entire dynamic range of fusion power. This setup also provides sensitivity to different energy groups: 235U is sensitive to neutrons in a wide energy range, especially at low energies, while 238U is sensitive to neutrons above the threshold of ~ 1 MeV.
ITER in situ calibrations are planned prior to the Start of Research Operation (SRO) and the DT-1 operational phase. NG-24M sealed tube neutron generator will be mounted on a robotic arm deployed inside the vacuum vessel from the Equatorial Ports. The in situ calibration campaigns aim to determine the calibration coefficients of the DNFM fission chambers that convert the detector count rates into total neutron yield. The presence of a robotic arm in the vacuum vessel introduces a bias in the calibration measurements due to neutron scattering on its structure. One of the objectives of this work is to quantify this source of bias. Monte Carlo neutron transport simulations employing a detailed model of the ITER machine, material compositions, and neutron source characteristics are used to calculate the uranium fission rate in DNFM fission chambers.
In this work, the OpenMC code is used to assess the impact of the robotic arm on the detector count-rates. In four of the six cases considered, the presence of the robotic arm alters the fission rates by up to 13%, with the largest contribution originating from the redistribution of neutrons below 1 MeV. We further demonstrate that positioning the neutron generator closer to the detector and away from the plasma axis leads to a neutron spectrum profile that is significantly closer to that of the volumetric plasma source.
The assessment underlines the need for detailed geometric modelling of both the in situ calibration campaign and of the machine during operation. This work contributes to the determination of the DNFM calibration coefficients and the associated uncertainties. Neutron transport simulations performed with the OpenMC code enable this bias to be explicitly evaluated and incorporated into the calibration coefficients, thereby compensating for its influence. The obtained results allows us to formulate a set of proposed calibration source positions.
The work is supported with the task agreement between the Institution “Project Center ITER” and the ITER International Fusion Energy Organization “Neutron calibration equipment design, modelling and testing” (IO/25/TA/450000025) as of 12th of December 2025.
The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

Speaker: Egor Afanasenko (ITER RF DA)

fnm-26-abstract-eng-final-2.docx

10:40 → 11:10 Coffee Break

11:10 → 12:30 Fusion Reactor Design and Safety

Fusion reactor design and safety
Facility nuclear design challenges
Radiation mapping
Breeding blanket optimisation
Activation-related issues
Occupational radiation exposure and ALARA
Radioactive waste management

11:10 Stellarator reactor modeling and simulation approach using a fast-meshing method for MCNP6

The stellarator has re-emerged as a promising fusion concept, supported by the success of Wendelstein 7-X. Within EUROfusion, the HELIAS stellarator concept has previously been continuously developed. In addition, stellarator-based reactor concepts are increasingly emerging as a mainstream reactor concepts among private fusion companies worldwide, due to their grid-friendly operation and inherent plasma stability. However, the complex geometry of stellarators poses significant challenges for modeling and neutronics simulations. Existing approaches based on constructive solid geometry (CSG) and facet modelling still do not fully address those limitations in a streamlined analysis including fast in geometry modeling and integration, computational efficiency, and the application of variance reduction techniques.
At Karlsruhe Institute of Technology, a mesh-based modeling approach has been developed to enable rapid conversion from CAD models to unstructured meshes for neutronics simulations using MCNP6. The workflow starts with a CAD model of the stellarator, which is exported as a faceted STL geometry and subsequently tetrahedralized using the TetGen tool. This approach preserves boundary detail while reducing internal mesh complexity, allowing complete mesh generation of complex stellarator geometries within minutes. The method is integrated with tools such as SpaceClaim and the McCad–SALOME platform to provide a streamlined pipeline for facet export, mesh generation, and MCNP6 input preparation. The approach has been validated using an open stellarator model available on GitHub. Benchmarking models against established CSG-based models built with the SHANE and HeliasGeom tool for the new-generation optimized quasi-isodynamic configuration CIEMAT-QI4X has been performed. Comparisons of mesh tallies, cell tallies, and nuclear responses demonstrate the accuracy and efficiency of the proposed method.

Speaker: Yuefeng Qiu (KIT)

YQU_mesh_approach_for_Stellarator_modelling_v0.4.docx

11:30 Assessment of Divertor Activation Caused by 14.1 MeV Neutron Streaming through Monoblock and Inter-Module Gaps

The divertor is an essential component in fusion reactors and must operate under extreme thermal loads and intense particle flux conditions. Consequently, most divertor designs, including ITER, adopt a water-cooled tungsten (W) monoblock (MB) concept. For fusion devices currently under construction or planned, neutrons generated by fusion reactions represent a critical design factor alongside high heat load. Thus, the divertor must play a vital role in shielding the vacuum vessel and superconducting magnets. From this perspective, neutron streaming occurring through inherent geometric gaps such as the poloidal gaps between MBs and the toroidal gaps between divertor modules is a crucial point in the fusion devices considering burning plasma. These gap regions are not protected by W armor, leading to direct exposure of divertor components and the vacuum vessel to high energy neutrons. As a result, neutron flux in these regions is higher than in W shielded areas, causing increased neutron-induced damage such as displacements per atom and helium production. In addition, these conditions significantly affect material activation due to the high-energy neutrons of 14.1 MeV.
In this study, the impact of neutron streaming through MB and inter-module gaps on divertor activation was analyzed. A simplified divertor model based on the Compact Pilot Device (CPD), currently under conceptual design in Korea, was employed. The model consists of W armor, a copper (Cu) interlayer, a CuCrZr heat sink, water coolant, and a cassette body made of Advanced Reduced Activation Alloy (ARAA). Compared to ITER, ARAA, a reduced-activation ferritic/martensitic steel, is used for the cassette body to mitigate long-term activation. Neutron transport calculations were performed using MCNP6 with the FENDL-3.1 library, and activation analyses were conducted using FISPACT-II with the TENDL-2017 library.
The results indicate that neutron flux in the gap regions of W, Cu, CuCrZr, and the ARAA cassette body is approximately 6–11% higher than that beneath the MBs. When this flux distribution is applied to activation calculations over a 100-year cooling period, the activation level in gap regions is found to be 11–37% higher than in MB regions. The surface of the cassette body shows differences of up to 65%. These results demonstrate that exposure to high energy neutrons through gaps can significantly amplify activation, even for relatively small increases in neutron flux.

Speaker: Seonghee Hong (Korea Institute of Fusion Energy)

2nd_FNM_abstract_shhong_v1.1.docx

11:50 Radiation safety analyses of EAST tokamak for the D-T experiment campaign

With the upgrade of in-vessel components, heating systems and the construction of specialized tritium storage, injection and collection system, EAST Tokamak will be ready for a short D-T operation phase (weeks) embedded within longer D-D phases, by the end of 2026 or later. This report summarizes the radiation safety analyses conducted for this campaign, to address the satisfaction of radiation safety requirements approved by the authority body.

The characterized radiation source terms are the prompt radiation, induced radioactivity in materials, air and water, the tritium. The main measures for radiation safety include the functioning of the existing tokamak hall with 1.5m-thick concrete walls, 1m-thick concrete roof and radiation shielding doors, the implementation of a radiation zoning and access control, and the upgraded radiation safety monitoring and interlock system. In addition, specific tritium safety measures, including a multi-layer tritium containment system (vacuum boundary, gloveboxes, sealed hall) will be implemented. The analyses indicated that radiation safety risks will be well-controlled. The dose rates outside the hall are below the control level of 2.5 µSv/h during operation and the shutdown dose rates inside the hall will below the control level in an acceptable time. The integrate dose of the radiation workers and the public are significantly below regulatory constraints of 5 mSv/a and 0.1 mSv/a, respectively. Radioactive waste management ensures no planned release of tritium, gaseous and liquid effluents are captured and stored, while solid waste is stored or disposed of via authorized routes. Preliminary accident analyses demonstrate that the dose of the public remains below the dose constraint 1 mSv even under severe external events.

The analyses confirm that adequate protection for radiation workers, the public, and the environment is assured with the technical and administrative measures for the EAST D-T campaign.

Speaker: Kun XU (ASIPP)

Radiation safety analyses of EAST tokamak for the D-T experiment campaign_abstract_V1.0_ASIPP-KunXU_2026-04-10.docx

12:10 OpenMC Verification and Validation for Well Logging Applications

Content in attached pdf.

Speaker: Bor Kos (Baker Hughes)

OpenMC-WellLogging-KosVogtBanzarov-FNM-v1.0.pdf

12:30 → 13:00 Closing Session

13:00 → 14:00 Lunch Time

14:00 → 16:00 Technical visits (Friday, June 12th, full afternoon)

14:00 KIT Fusion Materials Lab (FML)

14:00 KIT Helium Loop (HELOKA-HP)

14:00 KIT Research Accelerator (KARA)

14:00 KIT Tritium Laboratory (TLK)