Conference Agenda

Since 1984, the number of occupational radiation workers in Korea has steadily increased, now almost approaching 50,000 individuals. Globally, approximately 24 million workers are registered as occupationally exposed to radiation. As this population grows, the demand for fast, large-scale dose assessment methods with sufficient accuracy and reliability becomes increasingly important for ensuring the safety and protection of radiation workers. Traditionally, decision tree (DT) based algorithms have been utilized for evaluating doses using thermoluminescence dosimeters (TLDs). Despite its wide usage, DT methods are inherently time-consuming and labor-intensive to develop. Also, they exhibit some critical limitations, such as reduced classification accuracy for high-voltage X-ray fields and difficulties in accurately assessing the mixing ratios in mixed radiation-field environments.

Previous studies have explored the application of artificial neural network (ANN) techniques for TLD data analysis, reporting modest improvements in dose prediction accuracy over conventional DT approaches. However, these models often failed to identify high-energy photon sources such as pure Cs-137, leaving them inadequate for comprehensive dose assessment across all radiation field types [1]. To address such issues, one past study demonstrated an alternative ANN-based unfolding method to reconstruct the energy spectrum of the incident radiation field using multi-element data from optically stimulated luminescent dosimeters (OSLDs) [2]. By estimating the spectral distribution, the method enabled the calculation of energy-specific fluxes and corresponding effective doses by applying appropriate dose conversion coefficients. Despite its improved field characterization capability, this approach is constrained by its complexity, requiring extensive spectral data for model training and the derivation of kerma response functions. Moreover, it remains inapplicable to beta radiation fields.

To overcome the limitations of earlier methods, we developed a dose assessment algorithm for TLDs that exhibits higher accuracy and computational time by leveraging machine learning techniques. In particular, the algorithm is based on the Light Gradient Boosting Machine (LightGBM) framework that focuses on the radiation field classification, performance quotient evaluation, and H_p(0.07) and H_p(10). The training and testing datasets were generated by augmenting the reference element response data from LiF:Mg,Cu,Na,Si TL material, developed by KAERI, with its high reproducibility of within 3%. We evaluated the performance of the LightGBM algorithm in comparison to ANN and the traditional DT method. Each algorithm comprises five sub-models that categorize radiation fields and quantify mixed-field ratios. Among them, the LightGBM model demonstrated superior performance in radiation field classification and mixing fraction estimation with the lowest performance quotients.

Furthermore, when comparing the deviations between predicted and actual shallow and deep dose equivalents, LGBM consistently exhibited the smallest discrepancies. This suggests that LGBM yields reduced bias and standard deviation, enhancing the overall accuracy and precision of dose assessments compared to the DT model. Furthermore, it achieved the highest accuracy in dose assessment among the tested models, establishing it as a viable alternative for broader implementation in radiation protection practices.

[1] H.Y. Jung, Development of personal radiation dosimeter and dose assessment algorithm based on LiF thermoluminescence material and artificial neural network, Ph.D. dissertation, KAIST, 2003.

[2] S.-Y. Lee, B.-H. Kim, K.J. Lee, An application of artificial neural intelligence for personal dose assessment using a multi-area OSL dosimetry system, Radiation Measurements 33(3) (2001) 293-304. https://doi.org/https://doi.org/10.1016/S1350-4487(00)00147-5.

Rapid, reliable localisation of several radioactive sources is a cornerstone of modern emergency response, decommissioning, and safeguards inspection, yet today’s survey methods still falter whenever two or more emitters lie close together or possess similar intensities. We report a fully deployable, network-enabled radiation-mapping platform that overcomes these limitations by combining a novel probe architecture, swarming logistics based on unmanned aerial vehicles, and an inference engine that turns simple shielding effects into high-confidence directional clues. The concept was first conceived as a lightweight upgrade for existing drone-borne monitoring teams, but has since matured into a self-contained system that can be scattered over unknown terrain, autonomously build a gamma-ray map, and return precise source coordinates within minutes—without any human entering the hot zone.

Each probe, weighing roughly four hundred grams, integrates three key elements. A main poly-vinyl toluene scintillator counts incoming gamma quanta, while at least one auxiliary scintillator is mounted on the opposite side of a purpose-built shield. The shield introduces a sharply defined angular sector—labelled “alpha” in our internal documentation—in which photons are preferentially attenuated before reaching the main detector. When a source happens to lie inside this shaded window the auxiliary detector registers a noticeably higher rate than the main one; if the source lies elsewhere the rates remain comparable. To interpret this contrast after the probes tumble to the ground, miniature gyroscopes and accelerometers log the final attitude, and a global positioning receiver tags every record with absolute latitude, longitude, and altitude. A miniature CsI(Tl) spectrometer converts pulse-height distributions into digital form, making it possible to preserve isotopic fingerprints for downstream analysis.

The probes are carried to the scene by drones or other robotic carriers. During flight a low-power mesh radio link is established through airborne relay nodes called “agents”; upon release the probes parachute or free-fall, start counting immediately, and continue transmitting raw spectral data, orientation vectors, and position coordinates to a ground server through the mesh. Because the network re-routes automatically around obstacles or dropouts, coverage is maintained even if some relays are lost.

Source localisation proceeds in two logical passes executed by the server. First, for every probe the algorithm compares the count rates of the paired detectors. A pronounced inversion—auxiliary exceeding main—flags the probability that a source lies within the probe’s alpha sector; a flat ratio implies it lies in the complementary region that we denote “beta.” This decision yields a probe-wise directional estimate rather than a single line of bearing, effectively casting a wide yet bounded fan in the suspected direction. Second, the server overlays all directional fans from the swarm. The area where they overlap most strongly contracts rapidly as more probes report, often shrinking to a patch only a few metres across in less than one hundred seconds. By relying on overlaps rather than on geometric back-projection with narrow collimators, the method remains tolerant of background fluctuations and of modest errors in the inertial sensors.

Several design choices directly lower occupational dose. Because directional information is extracted from the built-in shield rather than from a heavy tungsten collimator, each unit stays light enough for drones to carry dozens per sortie, and field teams can remain outside the hazard boundary until the map is complete. The platform is inherently scalable; adding extra probes or additional drone passes simply supplies more directional votes, further tightening the uncertainty without changing the underlying logic. Modularity also allows future substitution of neutron-sensitive plastics, solid-state cadmium-zinc-telluride detectors, or advanced digital processors.

Current development focuses on coupling the drones’ own flight tracks to the localisation solver, refining the Bayesian fusion layer that already screens pulse-height spectra for isotope hints, and conducting extended range exercises with independent evaluators. In summary, by turning a simple shielding asymmetry into a powerful directional discriminator and networking many small sensors, the deployable platform delivers a practical, rapid, and safer path to surveying complex radiological scenes that outperforms legacy methods in both speed and accuracy.

In the event of a severe nuclear accident, a prolonged interruption in the supply of cooling water to the reactor core can result in excessive heating of the fuel elements, ultimately compromising the structural integrity of the core. In such scenarios, the nuclear fuel begins to melt due to its own decay heat and subsequently interacts with surrounding structural materials, leading to the formation of a complex, porous, and heat-generating structure referred to as a debris bed. Under extreme conditions, the molten core material may relocate to the lower plenum of the reactor pressure vessel (RPV). In the presence of residual cooling water, the material may fragment and form a debris bed composed of irregularly shaped and sized particles. To prevent the remelting of this debris and the subsequent structural failure of the RPV, it is crucial to ensure effective quenching of the debris bed. Inadequate quenching can result in vessel failure and the uncontrolled release of molten core material, posing significant safety risks. [1][2]

Understanding and optimizing the coolability of the debris bed necessitates careful consideration of various structural and operational parameters, including porosity, particle geometry, and the flooding approach. [3] Consequently, experimental investigations play a critical role in studying the quenching behaviour of debris beds under controlled conditions. One such experimental facility is the FLOAT facility at the Institute of Nuclear Technology and Energy Systems (IKE), University of Stuttgart. The FLOAT facility features a cylindrical debris bed with a diameter of 200 mm and a height of 300 mm, which is heated using an inductive heating system to achieve the required initial temperature distribution. Quenching experiments are performed under a range of conditions, systematically varying parameters such as initial bed temperature, porosity, and particle arrangement. Although comprehensive temperature data are obtained from 45 thermocouples embedded within the bed and near the wall, detailed visual data characterizing the two-phase flow near the wall have not yet been analysed. To address this limitation, an array of 12 video cameras is positioned around the transparent glass outer wall of the test section. These cameras capture high-resolution images during the quenching process, enabling the visualization of quench front progression and two-phase flow behaviour within the debris bed.

The recorded images undergo a series of pre-processing steps, including reflection removal, noise reduction, and normalization, to enhance the visibility of relevant features distinguishing wet from dry regions. Following pre-processing, the images are manually labelled to create a high-quality labelled dataset that identifies wet and dry regions within the debris bed. This labelled dataset serves as the basis for training machine learning models aimed at automating the analysis of the experimental imagery. A Convolutional Neural Network (CNN) architecture is selected due to its particular efficacy in recognizing complex visual patterns and performing classifications on image data, making them an optimal choice for this application.

The CNN model is trained using the labelled dataset, enabling it to learn to distinguish between wet and dry regions based on the extracted image features. During training, model performance is evaluated using a validation dataset to ensure appropriate generalization and to prevent overfitting. Upon completion of training, the model is tested using a set of new images to rigorously assess its generalization capability. Finally, accuracy, precision, metrics which are used to evaluate the performance of classification models e.g., F1-score, and Structural Similarity Index Measure (SSIM) of the model are assessed to ensure reliable performance.

This integrated approach, combining advanced experimental techniques and machine learning methodologies, facilitates the automated and accurate identification of two-phase flow patterns and quench front progression within the debris bed. The resulting insights significantly contribute to the broader understanding of debris bed coolability and support the development of improved safety strategies for severe accident management in nuclear reactors.

References:

[1] Ahmed, Z. et al. (2022) ‘Experimental investigation on the coolability of nuclear reactor debris beds using seawater’, International Journal of Heat and Mass Transfer, 184, p. 122347. doi:10.1016/j.ijheatmasstransfer.2021.122347.

[2] Leininger, S., Kulenovic, R., and Laurien, E. (2015) ‘Experimental Investigations on the Coolability of Debris Beds under Variation of Inflow Conditions’, Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16, Chicago, IL, August 30–September 4, 2015. Available at: https://glc.ans.org/nureth-16/data/papers/12903.pdf.

[3] Li, Y. et al. (2025) ‘Debris Bed coolability in hypothetical core disruptive accidents: Theory, experiment, and Numerical Simulation Review’, Progress in Nuclear Energy, 184, p. 105700. doi:10.1016/j.pnucene.2025.105700.

Background:

Effective learning from past incidents is essential to improve the safe and reliable operation of nuclear power plants (NPPs). Operating-experience databases contain valuable knowledge, but traditional search systems often rely on literal keyword matches or manual classifications, making it difficult to retrieve relevant information from large document collections. This limits their capacity to support operational decision-making and safety improvements.
Objectives and Approach:

Project CERO, supported by the Spanish Nuclear Safety Council and based on the NRC’s Agencywide Documents Access and Management System (ADAMS) database, addresses this challenge by applying the latest advances in deep learning and natural language processing (NLP) to the management of operating experience. We have developed an AI-powered search and analysis tool capable of retrieving, organizing, and connecting incidents in a way that supports learning and proactive safety strategies.
Methods:

The system we have developed builds upon three main pillars: a unified metadata schema for standardized report information, automatic extraction and ranking of keyphrases, and a semantic-search model based on vector embeddings. Incident reports are processed through a two-step pipeline. In the offline phase, 5375 Licensee Event Reports (LERs) from 2004 to 2024 were collected and processed. Using a TF-IDF-based ranking method combined with syntactic filters, keyphrase candidates were identified from each incident’s abstract and narrative sections. A disambiguation step, based on AI techniques including semantic embeddings and edit distance, was applied to unify variants of the same concept, correcting some of the errors introduced by the optical character recognition (OCR) process used on the scanned documents.

In total, 52623 keyphrases were extracted across the dataset, covering multiple sections such as Title, Causes, Initial Conditions, and Corrective Actions. Approximately 6.8% of the candidate keyphrases were successfully disambiguated. The top-ranked keyphrases were embedded using a multilingual deep learning model, enabling semantic search capabilities, including support for cross-lingual queries beyond English.

In the online phase, when a user submits a query, it is vectorized using the same deep learning embedding model. Keyphrases are retrieved based on semantic similarity, and associated incident reports are ranked according to a weighted combination of similarity scores and keyphrase importance. An approximate nearest neighbor (ANN) search structure ensures the system remains scalable while maintaining high retrieval precision. Compared to traditional literal-match systems, our AI-powered engine consistently returns a greater number of relevant results, and qualitative testing indicates that the results better align with user intent and operational needs.

Beyond simple retrieval, we apply AI-driven semantic embeddings to construct a similarity-based network of incidents. Rather than relying solely on shared keyphrases, we compute embeddings for the full incident narratives and link documents based on vector similarity. This embedding-based approach produces a consistent and coherent network of related incidents. This network is being analyzed using complex network analysis techniques, such as community detection algorithms, to identify clusters of recurring issues, common causes, and systemic vulnerabilities. Early results show that clusters often correspond to known categories of technical or operational concern, validating the potential of this approach to support proactive safety management.
Impact and Future Work:

The combination of semantic search, metadata unification, and network-based analysis represents a significant advancement in the use of AI for operating experience management. Our system not only facilitates faster and more intuitive access to relevant historical events but also provides a foundation for identifying patterns that would be difficult to uncover manually. Looking forward, the same framework can be extended to other incident reporting databases, both within the nuclear industry and across other high-reliability sectors.

We present the system’s architecture, preliminary results, and discuss its expected impact on improving operational safety, knowledge management, and incident-prevention strategies in nuclear power plants.

The increasing availability of large and complex datasets in the field of radiation protection opens new opportunities for the application of big data analytics. This study aims to explore how advanced data analysis techniques—including data mining, machine learning, and artificial intelligence—can support radiation protection goals by extracting meaningful insights from monitoring data. As a first step, we identify and characterize relevant datasets generated by decades of environmental and personal dosimetry monitoring, such as data from site-based gamma dose rate monitoring networks, meteorological stations at the Hungarian Paks Nuclear Power Plant, and Pille dosimetry measurements from the International Space Station. We review the current state of big data applications in radiation protection and propose a framework for future analyses. Preliminary work includes mapping data sources and outlining methodological approaches suitable for predictive modeling and anomaly detection. The results of this preparatory study will serve as a foundation for future work aiming to integrate big data methodologies into operational radiation protection.

In recent years, the nuclear engineering community has witnessed the emergence of several specialized open-source tools built upon the OpenFOAM framework, responding to growing demands for flexible and extensible simulation environments. Notable developments include GeN-Foam for coupled neutronics/thermal-hydraulics analysis, containmentFOAM for severe accident simulations, and OFFBEAT (OpenFOAM Fuel BEhavior Analysis Tool) for high-fidelity nuclear fuel modeling.

OFFBEAT distinguishes itself through multi-dimensional (1D, 2D, and 3D) simulation capabilities that capture complex phenomena such as pellet-cladding interaction, thermo-mechanical deformation, and fission gas release with high fidelity. By leveraging OpenFOAM's multiphysics architecture, OFFBEAT overcomes the limitations of traditional 1.5D fuel performance codes, offering crucial insights into localized phenomena like pellet-pellet contact and cladding deformation, factors critical for mechanical integrity and nuclear safety assessments. However, these high-resolution simulations demand substantial computational resources, necessitating efficient utilization of high-performance computing (HPC) infrastructure.

This study systematically evaluates OFFBEAT's HPC portability and parallel performance, examining how simulation dimensionality affects computational efficiency and demonstrating the scalability of high-fidelity nuclear fuel simulations on the CRESCO-HPC supercomputer. Our investigation compares two domain decomposition strategies native to OpenFOAM: the "simple" approach and a "manual" method, for which we developed a specialized mapping tool optimized for cylindrical fuel rod geometries.

To validate our approach, we conduct a comprehensive benchmark study comparing both serial and parallel OFFBEAT simulations against traditional 1.5D approaches. Results are presented in terms of output accuracy, computational efficiency, and resource utilization, quantifying the benefits and trade-offs between high-fidelity modeling and simplified approaches.

This work establishes a foundation for future large-scale, three-dimensional fuel rod simulations with OFFBEAT while offering relevant insights for the broader community of OpenFOAM-based nuclear applications. Since these codes share the same underlying computational framework, the parallelization challenges and optimization strategies identified here may benefit the entire ecosystem of OpenFOAM nuclear simulation tools. Ultimately, this analysis contributes to improving the practical applicability of advanced fuel performance codes in nuclear engineering research and industry applications.

The paper will present the content, experiences, good practices and feedbacks from radiation protection courses, organized at Nuclear Training Centre Ljubljana. Nuclear Training Centre is a part of Jožef Stefan Institute which is an authorised institution in the field of Radiation protection and Radiation protection training in Slovenia. A wide spectrum of courses for different users are regularly organized. The content of courses is different for various type of exposed workers and the courses are divided to three mayor fields: Radiation protection for industrial and other practices, Radiation protection for medicine and veterinary medicine and Radiation protection for nuclear facilities. Participants are requested to answer evaluation questionnaire at the end of all courses. All comments are then distributed to lecturers and the plan of necessary improvements is made.

The CONNECT-NM partnership is a pivotal initiative aimed at transforming nuclear materials research from an "observe and qualify" paradigm to a "design and control" approach. This shift is essential for enhancing the safety, efficiency, and sustainability of nuclear energy systems. Within this framework, the European Nuclear Education Network (ENEN) plays a critical role in ensuring the long-term impact of research through structured Education & Training (E&T) programs, mobility opportunities, and professional development initiatives. This paper explores ENEN’s contributions to CONNECT-NM, highlighting its efforts in disseminating research opportunities, fostering workforce development, and aligning nuclear materials research with industry and regulatory needs. A central element of ENEN's involvement is the establishment of a young researcher network, designed to foster collaboration and knowledge exchange among emerging professionals in nuclear materials science. Activities of the project such as the CONNECT-NM Brokerage Event, seek to facilitate collaboration between researchers and institutions to maximize project synergies. By integrating cutting-edge research with comprehensive training, ENEN contributes to it’s mission to strengthens Europe’s nuclear materials expertise, ensuring a robust and well-equipped next-generation workforce.

The Information Centre is part of the Nuclear Training Centre at the Jožef Stefan Institute and informs the visitors about nuclear power and nuclear technology, about radioactivity, about Krško Nuclear Power Plant and about energy in general.

Our main target population are the schoolchildren from the last grades of elementary school and from high school (ages 13-18), some 7000 visitors per year. We offer live lectures on nuclear technologies (fission and fusion), a lecture about use of radiation in medicine, industry and science and a lecture on stable isotopes. A general lecture about energy and an energy workshop is also available and usually performed for younger visitors. The visit includes a demonstration of radioactivity, a tour of our permanent exhibition and an optional tour of the TRIGA research reactor.

Since 1993 we monitor the opinion trends by polling some 1000 youngsters by 10 questions that remain unchanged for several years. This enables us to follow the trends in the basic knowledge of energy issues among youngsters and their attitude towards nuclear energy.

Reliable electrical distribution is a critical safety pillar of any nuclear power plant (NPP), ensuring that essential systems remain functional during both normal operations and emergencies. Among the most significant electrical faults is the Loss of Offsite Power (LOOP), a scenario in which the plant’s connection to the grid is suddenly lost. While LOOP is not in itself a severe accident, it is recognized in safety analyses as a high-risk initiating event that can escalate to a station blackout if emergency power systems fail.

This paper presents a two-day simulator-based training program conducted by the Philippine Nuclear Research Institute (PNRI) to strengthen competencies in managing LOOP events. Using a Generic Pressurized Water Reactor (GPWR) simulator, participants learned to observe, interpret, and analyze the plant’s real-time response to a LOOP, including the automatic startup of Emergency Diesel Generators (EDGs) and transfer of critical loads to the Class 1E safety-related network. The training emphasized not only technical system recovery but also the importance of early detection, situational awareness, and verification of safety functions.

By recreating the dynamics of a LOOP in a realistic yet risk-free environment, the program allowed participants to gain insights into accident progression, system interdependencies, and the critical role of electrical resilience in severe accident prevention. The results highlight the value of targeted, scenario-based training in reinforcing both technical proficiency and safety culture, ultimately supporting the safe and reliable operation of NPPs.

The geopolitical crisis in Ukraine has significantly impacted the national nuclear education and research landscape, posing serious challenges to both workforce continuity and integration with European structures. In response to this challenge, the NURECAB project (EU-UA Nuclear Research and Education Capacity Building, GA No. 101173510) was launched in 2024 under the EURATOM programme to modernize Ukrainian nuclear education and research and to facilitate its integration into the European nuclear research and academic area.

The core objective of the project is to close the gap between Ukrainian academic education and training and the evolving needs of the nuclear industry, particularly in the context of planned infrastructure expansion and new reactor deployment (e.g., AP1000, SMRs). A key instrument in achieving this is a comprehensive survey-based gap analysis involving over 49 Ukrainian institutions and 86 respondents from universities, research institutes, regulatory bodies, the national operator of nuclear plants, nuclear power plants staff, and other related stakeholders. Results have revealed persistent mismatches in curricula content, limited availability of the hands-on training, and insufficient coverage of trend reactor technologies and safety assessment practices.

Based on these insights, NURECAB has launched the development and revision of ten nuclear education courses, incorporating practical training modules and alignment with European Qualification Frameworks. Additionally, the integration of the e-learning platform ("ASKO") which is being used by the national nuclear power plants operator into university programmes is enabling more direct simulation-based learning. To build institutional capacity, the project implements training sessions for Ukrainian educators and researchers which cover modern reactor systems, fusion physics, up-to-date pedagogical approaches and safety culture. These include trainings for the Euratom NCP staff on EU project participation and policy frameworks.

Addressing the urgent need to attract and retain young professionals, the project has established a mobility grant scheme to support Ukrainian students and researchers in accessing international training, events and internship programmes. Targeted outreach campaigns have also been launched in cooperation with Ukrainian and European Nuclear Society, including national and international contests, career webinars and local engagement with the secondary school students.

The European Nuclear Education Network (ENEN) and Euratom National Contact Point in Ukraine (Euratom NCP) being leading EU partner and Ukrainian partner in the NURECAB project play a crucial role in linking Ukrainian institutions with the European community. ENEN has facilitated curriculum alignment, quality assurance and institutional integration, while enabling access to its extensive network of universities and research institutions and other education and training providers. The NURECAB project has established and successfully held a number of integration workshops in person and online for various target audience resulted in tangible integration outcomes, including the formal accession of some of the Ukrainian partners to the ENEN network.

The NURECAB project exemplifies how structured collaboration between EU and non-EU partners can uphold and advance nuclear research and education during crisis times through policy-informed, practice-oriented tools. It strengthens institutional links, modernizes training pathways and reinforces European collective nuclear knowledge base. This research explores the role of European Associations like ENEN in nuclear education and research integration and offers case examples of the models for building the nuclear competence under crisis conditions.

Keywords: nuclear education, capacity building, Ukraine, ENEN, Euratom, Euratom NCP, workforce development, nuclear competence, international cooperation, SMR, AP1000, training modernization

The construction of a new nuclear power plant (NPP) in Slovenia requires extensive early stage preparation, with workforce training being a strategic priority. Although the final decision has not yet been made, GEN energija, the company behind the Slovenia's new build, approached the Nuclear Training Centre (NTC) at the Jožef Stefan Institute to develop a dedicated training program for the preparation phase of the project, given that the center provides all theoretical training for the Krško NPP technical staff.

The initial consultations with GEN energija, identifying their needs, resulted in a list of topics, ranging from siting of a nuclear facility, characteristics of potential generation III NPPs to safety culture and leadership in nuclear facilities. Overall 20 topics are covered. NTC relied on its staff and external experts to finalize the curricula and draft training material, which took several phases of development, the final step being conduct of a pilot training course called New Nuclear Power Plants. The future actions include taking into account the feedback received from the pilot course and setting up the long term training plan for the GEN energija staff, as well as for professionals in the nuclear and regulatory sector.

The paper sheds light on challenges faced during the development, outlines the program and gives actual data on the pilot delivery.

In 2022, based on the legal requirement [1], the Slovenian Nuclear Safety Administration commissioned authorized radiation and nuclear safety experts to prepare a justification assessment for consumer products that contain added radionuclides and are already in use, for which a justification assessment has never been made under Slovenian legislation [2].

Slovenian legislation [2] defines consumer products as a device or an item into which radionuclides (one or more) have deliberately been incorporated or were produced by activation, or a device or an item which generates ionising radiation. The purpose of this is to add certain properties to the objects in question, such as luminosity, strength, etc. Such items are freely available or accessible to individuals from the public without special supervision or regulatory control by the competent authority during and after sale. It should be emphasized that consumer products are not defined as radiation sources.

The main purpose of the project task was to identify such products and to assess the justification of their continued use, to warn the public about potential risks, to publish recommended methods of safe handling, either for personal use or as parts of collections. Additionally, disposal of such items was also considered, either due to discontinuation of use or where the use is not justified

Based on the results obtained, informative material was prepared, which is available in both electronic and printed form. Examples and descriptions are given for the individual groups of consumer products (i.e. products with tritium and radium, glass objects or ceramics using radioactive paint or glaze, throated mesh used in gas lamps, welding electrodes with thorium, miscellaneous – various consumer products like blankets, pillows, pendants, underwear, face masks), recommendations and warnings for use and disposal, as well as alternative non-radioactive solutions that can be used instead. Currently, intensive distribution of the printed brochure is underway to recognized entities such as research institutions, companies involved in radiation activities, fairs, faculties, some sections of the Chamber of Craft and Small Business of Slovenia etc. Informing the professionals and public is also taking place through presentations at various events, conferences, etc.

[1] Decree on Radiation Activities (Official Gazette RS, Nos. 19/18 and 6/24)

[2] Ionising Radiation Protection and Nuclear Safety Act (Official Gazette RS, Nos. 76/17, 26/19, 172/21 and 18/23 – ZDU-1O)

Slovenia is again participating in the European Partnership on Radioactive waste Management (EURAD 2, co-funded by the European Union under Grant Agreement n°101166718) to further implement a joint strategic programme of research & development, strategic studies and knowledge management activities at the European level, bringing together and complementing EU Member States’ programmes in order to ensure cutting edge knowledge creation and preservation in view of delivering safe, responsible and publicly acceptable solutions for the management of radioactive waste throughout all programme phases (from “cradle to grave”) across Europe now and in the future. Accordingly, Slovenia participates in 11 Work Packages (WPs) of the EURAD 2 out of 17, currently foreseen to span in the period from 2024 to 2029, divided into the first wave (2 years) and a possible extension for the programme’s second wave in 2026. The WPs in which Slovenian institutions are involved cover a broad spectrum of themes, including strategic planning, technological innovation, safety optimisation, stakeholder engagement, and long-term knowledge management:

• WP2 - Knowledge Management (KM) focuses on creating a long-term, sustainable knowledge management programme to maintain a competent workforce in radioactive waste management (RWM). KM supports knowledge transfer between Member States at various implementation stages and promotes a collaborative framework to maintain and update the EURAD Roadmap.
• WP3 - Alternative RWM Strategies (ASTRA) addresses the feasibility and challenges of alternative radwaste management (RWM) strategies, especially for small inventory member states. It addresses long-term storage beyond design lifetimes, explores deep borehole disposal, identifies stakeholder concerns, and investigates shared and innovative solutions for problematic or small-volume wastes. ASTRA supports the development of adaptable and context-specific solutions, especially valuable for smaller or less advanced national programmes, e.g. Slovenian RWM programme.
• WP4 - WM for SMRs and future fuels (FORSAFF) is dedicated to understanding and recommending policy, regulatory, and management strategies for emerging technologies, e.g. small modular reactors (SMRs) and new fuel types, along with waste characterisation and predisposal strategies. It also seeks to identify stakeholder concerns and promote effective communication, reflecting a commitment to proactively anticipate and guide future RWM challenges.
• WP5 - Innovative characterisation techniques for large volumes (ICARUS) supports the development and harmonisation of advanced techniques to characterise the physical, chemical, and radiological properties of large waste volumes, especially low- and intermediate-level mixed waste, improving safe waste management.
• WP6 - Sustainable treatment and immobilisation of challenging waste (STREAM) contributes to the development of sustainable treatment and conditioning methods for challenging waste streams using low-carbon materials. It also includes assessment of environmental and economic impacts and evaluates waste acceptance compliance.
• WP9 - Innovative and new container/canister materials under disposal fields conditions: manufacturing feasibility and improved durability (InCoManD) investigates the long-term behaviour and degradation mechanisms of conventional and novel HLW container materials under repository-relevant conditions, aiming to enhance their durability through both experimental testing and modelling.
• WP11 - Impact of climate change on nuclear waste management (CLIMATE) identifies knowledge gaps and proposes research needs related to the impacts of climate change on waste management facilities during their construction and operation phases, including site-specific risk assessments and evolving climate scenarios.
• WP13 - HLW repository optimisation including Closure (OPTI) seeks to establish a shared understanding of optimisation strategies for HLW repositories. It identifies technical, regulatory, and socio-ethical challenges and develops case studies to explore design improvements and potential future R&D needs.
• WP14 - Near-surface disposal optimisation based on knowledge and understanding (SUDOKU) enhances scientific understanding of the long-term performance of cementitious barriers and covers in near-surface facilities, especially in relation to degradation processes and their effect on radionuclide migration.
• WP15 - Next generation Digital Twins to support Optimisation, Construction and Operation of surface and subsurface radioactive waste management facilities (DITOCO2030) seeks to define a path toward integrating digital twin technologies into comprehensive platforms for optimising construction and operational decision-making across the lifecycle of RWM facilities.
• WP17 - Criticality Safety for Final Disposal (CSFD) aims to enhance the understanding and optimisation of criticality safety measures for the final disposal of fissile waste. It focuses on scenario development, model validation, assessment methodologies, and understanding potential consequences of post-closure criticality events.

Slovenia contributes to the strategic optimisation of both deep geological and near-surface disposal concepts, supporting safety, cost-efficiency, and environmental performance through research on system design, material degradation (including corrosion), and radionuclide migration. It engages in the development of low-carbon treatment technologies, long-term container durability studies, life cycle assessment, and regulatory readiness for new reactor types. With active involvement in stakeholder and civil society dialogue, and exploration of digital twin technologies for facility optimisation, Slovenia demonstrates a strong commitment to a safe, science-based, and socially responsive European RWM system. We will present the EURAD 2 and current status in the WPs where Slovenia is involved

The Central storage of the new Croatian Radioactive waste management center is intended for storing institutional radioactive waste. The institutional radioactive waste encompasses radioactive materials produced in institutes, hospitals, industry etc. Depending on the type of the sources and their activities, some of radioactivity sources will be stored in dedicated shielded containers, and some in metal drums. The sources will be located at different positions within the storage building. In this paper, dose rates within the Central storage are estimated using MCNP6 code. Sampling from multiple scattered sources of different estimated intensities are analyzed here. The contribution from each gamma source is considered first all together, then combined, and finally each source is sampled separately. Additionally, since used neutron sources will also be stored there, the neutron and related gamma doses will be calculated. The dose rates are tallied at locations close to the sources where personnel are expected to spend some time during usual activities such as visual inspection and measurements. Because the yellowcake could be located close to spent neutron sources, the potential of neutron multiplication in Uranium ore will also be investigated.

Hyperspectral (HS) imaging is the act of observing many thin frequency bands simultaneously. Different objects can have unique spectral signatures, enabling some materials to be identified from a high enough resolution HS image.

However, HS imaging usually requires a combination of either multiple cameras, or multiple filters, and often necessitates computationally expensive image stitching. A fast, simple method of producing these hyperspectral responses would be helpful for categorizing objects, identifying rust progression and many other useful metrics in many industries. Nuclear decommissioning often has need for identification of contaminants on surfaces that are not visible to the human eye. Such decommissioning also involves objects with complex shapes, and difficulties in access.

Herein, a novel method of spatio-spectral HS imaging is demonstrated using a linear variable filter (LVF) across a camera sensor with rotation around the camera's no-parallax point using a robotic manipulator arm. The robotic manipulator arm enables HS panorama to be captured at any orientation and location within the arm’s working envelope without the need for precise and time-consuming setup.

Light intensity values in each group of pixel columns cover specific frequency bands due to the LVF. Images are captured in sequence around the panorama cylinder as the arm rotates the camera in pixel width angle intervals to ensure the images align into a panorama without the need for stitching. The images are layered in incrementing wavelength bands into a hypercube during capture. The resulting hypercube can be sliced to observe the whole scene at specific frequencies, or single pixels spectra can be extracted.

The system utilises a robotic manipulator arm suspended from a gantry frame capturing wide angles in minutes. The resulting remote HS capture of scenes with identification of fluorescent powders scattered over curved metallic surfaces through their expected spectral peaks shows the systems effectiveness.

As pressurized water reactors (PWRs) age, many of them, including the Krško NPP, will enter the decommissioning phase within the next few decades. While fast neutron reactions significantly influence the embrittlement of large steel components, thermal reactions - in particular the radiative capture of ⁵⁹Co and ¹⁵¹Eu - will become considerably important for determining the radioactive inventory during the decommissioning phase. At the beginning of the 25th operating cycle of the Krško NPP, an Ex-Vessel Neutron Dosimetry (EVND) program was established to continuously monitor fast neutron fluence and estimate the displacements per atom (DPA) in the beltline region of the reactor pressure vessel. Dosimetry capsules and gradient chains were employed, measuring different reaction rates such as ⁵⁴Fe(n,p)⁵⁴Mn and ⁵⁸Ni(n,p)⁵⁸Co to estimate the fast neutron fluence and ⁵⁹Co(n,γ)⁶⁰Co to estimate the thermal neutron fluence. State-of-the-art Monte Carlo calculations were performed at the Jožef Stefan Institute (JSI) and compared with the EVND measurements. Although the reaction rates of ⁵⁴Fe(n,p)⁵⁴Mn and ⁵⁸Ni(n,p)⁵⁸Co, which are sensitive to the fast part of the neutron spectrum, showed good agreement with the measured data, the reaction rate for ⁵⁹Co(n,γ)⁶⁰Co, which is sensitive to thermal neutrons, was overestimated. To investigate this discrepancy, the backscattering of neutrons from the concrete was investigated, hydrogen and oxygen content in the bioshield was analyzed, and a sensitivity study was performed by varying the hydrogen and oxygen concentrations in the concrete surrounding reactor pressure vessel. These adjustments led to a reduced backscattering of thermal neutrons and thus to a decrease in the calculated ⁵⁹Co(n,γ)⁶⁰Co reaction rates. The paper discusses the effects of these adjustments on the activation and embrittlement of the reactor pressure vessel and concrete.

High level objective of decommissioning and radwaste management is to close the life cycle of nuclear facilities and promote sustainable solutions allowing resources preservation.

Today, more than 200 nuclear reactors are under decommissioning worldwide; In this context it is key to propose sustainable solutions with a circular economy logic to save repository capacities, reduce raw materials consumption (iron, coal, nickel, etc.) and avoid CO2 production to the extent reasonable. With close to 50 years of operational experience, Cyclife developed optimised solution to tackle those current and future needs.

The management of very large components is an issue faced by many nuclear operators. PWR Steam Generators belongs to this category of metallic materials that have very challenging size and weight (up to more than 20m length, 6m diameter and up to 450 tons per unit). Such components cannot be easily transferred to and qualified for disposal as one piece. In most cases size reduction, decontamination and other process steps.

Rip and Ship strategy for steam generators and other large components aims to offer a derisked approach raising on a robust process in dedicated nuclear licensed treatment facilities. With a specialised process, each step of the value chain of the treatment is optimised to provide an efficient solution that will reduce the quantity of residual waste and shorten the decommissioning timeline. After characterisation and obtention of licenses the large component is loaded on a truck or a boat for transport as one piece (preferred) or in sections. Once arrived at the treatment facility, it is placed in buffer storage awaiting the treatment slot. Disassembly/size reduction, decontamination and melting are then implemented. Secondary waste are conditioned and returned to the owner.

Rip & Ship is a strategy that for decades has been chosen by customers in several European countries including France, Germany and the UK. To date, hundreds of large components have been shipped for treatment at Cyclife Sweden, among them 30 PWR steam generators and Magnox boilers.

The intended presentation will provide an overview of the treatment concepts and facilities as well as case studies related to recently performed and ongoing treatment projects for contaminated large components.

The paper summarizes the results of non-destructive (radiography and computed tomography) and destructive (light optical and scanning electron microscopy) evaluation of operational degradation of welded joints of EWS pipelines. Microbial corrosion caused by manganese-oxidizing microorganisms was assumed as a potential degradation mechanism. Large corrosion caverns were confirmed under the inner surface of pipelines. MnO₂ globules found on inner surface near caverns suggests the considered mechanism of MOMO.

The paper presents an overview of Accident Tolerant Fuel (ATF) concepts developed for Light Water Reactors (LWRs), with a primary focus on modifications to the cladding and fuel pellet design. ATF fuel concepts are meant to withstand severe accidents (SA) the reactor core for a longer period of time than the current fuel system, while also maintaining or improving fuel performance during normal operation.

The overview will include developments of coated zirconium alloys, alternative cladding materials such as FeCrAl alloys, molybdenum-based multilayer designs, and SiC/SiC ceramic composites. The focus will be placed on their enhanced resistance to high-temperature steam oxidation, improved mechanical integrity during accident conditions, and reduced hydrogen generation rates.

The paper will also summarize improvements in fuel pellet designs, particularly doped UO₂ pellets and ceramic microcell fuel structures. Doped fuel pellets containing Cr₂O₃ and Al₂O₃-Cr₂O₃ additives enhance grain size, dimensional stability, fission gas retention and resistance to pellet cladding interaction (PCI). Ceramic microcell pellets, which incorporate oxide-based fission product traps, further enhance retention of volatile isotopes and reduce internal pressure during transient conditions.

The paper will also consider the implementation of ATF in the EU Taxonomy since fulfilling the taxonomy requirements is important both for increasing a project's acceptability and for improving access to more affordable financial resources.

The final goal of the paper is to assess the overall potential of ATF systems to improve reactor safety without compromising operational efficiency. Emphasis will be placed on accident-mitigation capabilities, oxidation resistance at high temperatures and mechanical resilience. The reviewed ATF concepts represent a viable path forward in enhancing the robustness of nuclear fuel against extreme accident scenarios while supporting long-term sustainability goals in the nuclear sector.

Crack growth due to fatigue damage can significantly impact the integrity of structural components in nuclear power plants. These components are made of ductile materials such as stainless steels, which can withstand rather large plastic strains, as opposed to brittle materials, to avoid sudden failures. Part of the energy from far field loads in ductile materials is dissipated through plastic flow (yielding and hardening) mostly concentrated in a plastic zone ahead of the crack tip, which is absent in brittle materials. This affects the fatigue crack growth considerably and it has to be accounted for in the crack-growth prediction methods and models. One of them is the phase-field method for fracture, which regularizes the Griffith's theory of fracture and has the advantage of being (crack) path independent. This paper extends within the Abaqus finite-element framework, a phase-field method model for fatigue crack growth from the literature, originally developed for brittle materials. The extension considers an in-house implementation of plasticity models with and without hardening. The simulation results with plasticity-enhanced model include crack-growth predictions and recovery of the experimentally observed Paris law, and are compared to the results with models for brittle materials. Furthermore, the results are also compared to those obtained with engineering methods.

Thermal radiation properties of nuclear fuel, especially at high temperatures, are relevant to safety analysis. For this purpose, the spectral emissivity of fuel should be determined well beyond the melting temperature. Emissivity is defined as the ratio of the emissive power of a body to the emissive power of a blackbody at the same temperature. The total hemispherical emissivity is relevant to determining the thermal radiation from a molten core. The spectral emissivity is also necessary for reliable pyrometric measurements of fuel temperature, especially close to phase transition. A review of experimental results published in the open literature has been performed. Recent results showed to be in agreement with recommendations provided by Bober and his co-authors.

Combined neutron and gamma irradiation induces significant alterations in the microstructure and mechanical performance of cement-based materials. At the bulk scale, concrete exposed to neutron irradiation exceeding 10¹⁹ n/cm² and gamma doses exceeding 200 MGy undergo progressive degradation, manifested by a reduction in compressive strength of up to 30 – 50%. This mechanical deterioration results from the synergistic effects of neutron radiation-induced volumetric expansion (RIVE) of crystalline aggregates and gamma radiation induced hydrolysis in cementitious phases, including calcium silicate hydrate (C–S–H) gel, portlandite (Ca(OH)₂), and residual clinker.

T assess the effects of radiation on hydrated cement paste, direct measurements of elasticity, tensile strength, and fracture energy at the microscale are important for investigation how the individual phases respond to gamma irradiation. This necessitates the fabrication of micro-beams using focused ion beam (FIB) milling, followed by mechanical testing through nanoindentation and micro-mechanical techniques.

Those measurements at the microscale can provide critical validation and calibration data for multi-scale modelling approaches, bridging the gap between molecular simulations and macroscopic behaviour. Understanding the phase-specific degradation mechanisms is essential for designing more radiation-resistant cementitious materials for nuclear infrastructure and long-term waste storage applications.

This work primarily focuses on the fabrication of FIB cantilever beams and their subsequent testing using nanoindentation bending, including the optimization of sample preparation and the influence of processing parameters on the mechanical properties of cement phases. The samples were subjected to varying doses of gamma irradiation to evaluate radiation-induced changes at the microscale.

Over the past several years, neutron irradiation of nanocrystalline anatase has emerged as a subject of considerable scientific interest, revealing notable structural and compositional transformations with potential implications for advanced functional materials [1, 2]. Anatase-phase TiO₂ nanoparticles were synthesized using standardized protocols to ensure uniform particle size and crystallinity. These nanoparticles were subsequently subjected to neutron irradiation with fluences ranging from 1.6 × 10¹⁵ n·cm⁻² to 2 × 10¹⁷ n·cm⁻², in order to investigate the effects of high-energy neutron exposure on their structural and surface properties. Fourier Transform Infrared (FTIR) spectroscopy was employed to acquire infrared absorption spectra both prior to and following irradiation, enabling detailed assessment of irradiation-induced structural modifications [3]. The FTIR spectra of the pristine (non-irradiated) anatase nanoparticles exhibited characteristic absorption bands attributed to Ti–O and Ti–O–Ti stretching vibrations, prominently centered around 450 cm⁻¹. These bands are indicative of the structural integrity and crystallinity of the anatase TiO₂ lattice. Upon neutron irradiation, notable spectral changes were observed, including the emergence of new peaks at 750 cm⁻¹ and 970 cm⁻¹, which correspond to deformation vibrations of the TiO₆ octahedral units. These features are interpreted as evidence of structural rearrangements within the anatase matrix, induced by neutron–matter interactions. The observed changes are attributed to energy transfer from neutron collisions, resulting in lattice distortions and local structural reconfigurations.

Neutron transmutation reactions occurring within the anatase (TiO₂) nanoparticles led to the formation of vanadium isotopes, as evidenced by the emergence of distinct infrared spectral features corresponding to V–O and V–O–H deformation vibrations. These newly formed peaks serve as clear indicators of vanadium incorporation into the TiO₂ lattice, thereby altering the chemical composition and potentially modifying the functional properties of the material. Additionally, increased surface activity was detected through the appearance of Ti–O–H deformation modes, suggesting that neutron irradiation facilitated the formation of reactive surface sites. This enhancement in surface reactivity is likely attributable to hydration processes and partial surface amorphization induced by radiation damage. The onset of amorphization was further confirmed by the broadening of absorption bands and the reduction in intensity of the characteristic Ti–O–Ti vibrational modes associated with the crystalline anatase structure. Despite these alterations, the persistence of the Ti–O–Ti peak at approximately 450 cm⁻¹ indicates that a fraction of the anatase lattice retained its crystallinity post-irradiation. Nonetheless, the presence of transmutation-related spectral features and vibrational shifts reflects significant structural reorganization. FTIR analysis thus provides compelling evidence of both structural and compositional transformations driven by neutron exposure. The observed formation of vanadium isotopes and the generation of reactive surface functionalities underscore the profound impact of neutron irradiation on nanocrystalline anatase. These findings have important implications for the design of radiation-tolerant materials and for catalytic systems where surface activity and structural adaptability are critical performance factors. Overall, the study highlights the potential of neutron irradiation as a tool for tailoring the physicochemical properties of nanostructured materials through controlled transmutation and defect engineering.

This article/publication is based upon the work from COST Action CA20129 MultIChem, supported by COST (European Cooperation in Science and Technology).

1. Maria Vasilopoulou et al. "Hydrogen and nitrogen codoping of anatase TiO2 for efficiency enhancement in organic solar cells" Scientific Reports 7, 17839, 2017
2. Elchin M. Huseynov, Efsane A. Huseynova, Anze Jazbec "EPR spectroscopic characterization of neutron-irradiated nanocrystalline anatase particles" Physica B: Condensed Matter 699, 2025, 416786
3. Elchin M. Huseynov, et al. "Infrared spectroscopy of nanocrystalline anatase (TiO₂) particles under the neutron irradiation" Optical Materials144, 114351, 2023

Additive manufacturing represents a cutting-edge technology that offers significant reductions in both manufacturing time and cost. However, any new technology or material needs to go through a qualification process, before it can be used in the nuclear industry. This paper reports on a pre-qualification process of additively manufactured 316H stainless steel. Laser Powder Bed Fusion (L-PBF) manufactured 316H in an as-built and solution-annealed state is compared with solution annealed wrought material. A Small Punch Creep test campaign at 650^oC was performed at different loads in inert (argon) environment. Since the campaign is still on-going, the initial Small Punch Creep results are reported, along with microstructural analysis of the as built L-PBF 316H ruptured test pieces. The L-PBF 316H results are compared with L-PBF 316L Small Punch Creep results obtained within the NUCOBAM EU-funded research project.

Population and life-standard growth increase demand in all sectors, including energy consumption. Because bigger overall consumption in an already changing climate due to human activities additionally contributes to the whole carbon footprint, population and life-standard growth only accelerate climate change in a negative and unpredictable way. Therefore, climate mitigating measures are investigated globally, at least in the technical part, also with introduction of new Small Modular Reactors (SMRs) for energy production, which are considered as one of the potential solutions in the energy sector. However, any new solution that might affect the population should be holistically analysed, public perception included, before any investments into real-life implementation are made.

While technical innovation offers promising pathways, its success ultimately depends on societal acceptance and alignment with public values. Nuclear technologies, including SMRs, often provoke complex societal responses, shaped by historical experience, perceived risks, and trust in institutions. As such, understanding how SMRs are viewed by both experts and the public is crucial for their responsible development and integration into national energy strategies.

Therefore, the public and expert perception of SMRs in Slovenia, currently pursuing a second nuclear power plant (NEK2) and evaluating SMRs as a potential future supplement to Slovenia’s energy mix, was investigated in the ECOSENS project (Economic and Social Considerations for the Future of Nuclear Energy in Society, funding from the Euratom Research and Training programme under grant agreement No 101060920). The study used qualitative methods, including semi-structured interviews with key national stakeholders and a focus group discussion with knowledgeable experts in energy sector. The research aimed to assess awareness, attitudes, and expectations regarding SMRs, particularly in relation to climate change mitigation, energy security, and national energy.

Findings revealed that while SMRs are mentioned in strategic documents such as the National Energy and Climate Plan (NEPN), they remain largely hypothetical, with no concrete deployment plans before 2050. Stakeholders agree that Slovenia is unlikely to pioneer SMR implementation but could adopt mature foreign technologies in the future. Investment uncertainties, lack of tailored legislation, and limited nuclear workforce capacity present key barriers. GEN energija, the main nuclear developer, has initiated exploratory efforts but lacks sufficient human resources to lead an SMR project independently.

The public discourse on SMRs is overshadowed by the more imminent NEK2 project. Most participants in the expert focus group were sceptical of SMRs’ feasibility, citing concerns about cost, regulatory readiness, waste management, and the real sustainability of “small” solutions. While some saw potential in decentralised generation and combined heat-power applications, others warned of repeating the shortcomings of larger nuclear projects under the guise of innovation.

Notably, the perception of SMRs is shaped not only by technical arguments but also by broader debates on nuclear energy’s role in Slovenia’s low-carbon future. Participants emphasised the need for transparent, inclusive public communication and rigorous decision-making frameworks. The experience with the Krško NPP has created a baseline of acceptance, but the introduction of SMRs would require a renewed social contract, especially given their proposed proximity to populated areas, also without nuclear experience.

The research on the Slovenian case underscores that SMRs are not just a technological proposition but a societal one. To ensure informed decisions and responsible innovation, SMR development must proceed alongside robust governance, stakeholder engagement, and critical reflection on energy futures.

GEN energija is actively exploring the deployment potential of Small Modular Reactors (SMRs) as part of its long-term decarbonisation strategy and the modernisation of its electric system. In light of increasing electricity consumption, driven by electrification trends and the need for stable, low-carbon energy generation, SMRs are considered a complementary solution to large-scale nuclear and renewable energy sources. This paper presents an overview of the current status of SMR-related activities in Slovenia, with a focus on the role of GEN energija and its ongoing SMR-related studies.

The starting point for considering SMRs in Slovenia is outlined in the country’s latest National Energy and Climate Plan (NEPN 2024), which explicitly includes the construction of one SMR unit by 2050, in parallel with the deployment of a larger nuclear unit by 2040. This marks a significant step in national policy, recognising SMRs as a potentially strategic asset for low-carbon electricity generation.

The key motivations for SMR consideration are threefold: (1) the need to diversify the low-carbon electricity mix and increase energy security; (2) the flexibility and modularity of SMRs, which allow for phased deployment and integration into various energy systems; and (3) the opportunity to repurpose or upgrade existing thermal power plant sites and infrastructure.

This paper presents an in-depth overview of the typology of SMRs being considered, and how these relate to Slovenia’s specific energy system and geographic conditions. Particular attention is given to the three major technology groups:

1. Light Water SMRs with Integral reactor – such as NuScale, i-SMR (KHNP), and CNNC’s ACP-100 design – where the entire primary circuit (reactor core, steam generators, pressuriser) is integrated into a single pressure vessel. These offer advantages in safety and compactness but raise challenges in long-term maintenance access and material ageing under higher neutron fluxes.

2. PWR SMRs with separated primary loop components, such as Holtec SMR-300 and Rolls-Royce SMR, which are technologically closer to existing large PWRs but apply modular construction and standardisation to reduce costs and construction time.

3. Advanced (GEN IV+) SMRs, including Natrium (sodium-cooled fast reactor with molten salt energy storage), X-energy’s Xe-100 (helium-cooled high-temperature reactor), and Newcleo’s LFR (lead-cooled fast reactor). These concepts offer enhanced efficiency, fuel cycle sustainability, and operational flexibility, but are at earlier stages of commercial maturity.

SMRs represent an innovative approach to addressing the well-known challenges associated with large-scale nuclear power plants. The modular construction of SMRs enables factory-based serial manufacturing and allows for accelerated and more predictable on-site assembly, thereby reducing project complexity and construction risks. Furthermore, the lower initial capital investment facilitates phased deployment, offering a flexible response to evolving energy system demands. SMRs provide versatility not only in electricity generation but also in supplying heat for industrial processes, district heating, and hydrogen production. Their compact physical footprint increases the range of viable locations, including the potential repurposing of existing fossil-fuel power plant sites, thus supporting the transition toward a low-carbon energy system.

Nevertheless, the pathway to the widespread deployment of SMRs is accompanied by significant challenges. Economic competitiveness remains uncertain, as the absence of large-scale serial production may result in higher levelized costs of electricity (LCOE) compared to conventional large reactors. Limited operational experience, particularly with integral and advanced reactor designs, raises valid concerns regarding maintenance practices and long-term reliability. Public acceptance will play a pivotal role, as nuclear technologies — regardless of size — require transparent communication and proactive public engagement to build trust. Additionally, supply chains for critical reactor components are not yet fully established, and licensing processes are still evolving. SMRs remain a promising technology for the near future, with the successful delivery of the first FOAK (First Of A Kind) projects in Europe or North America expected to play a decisive role in unlocking broader commercial deployment.

In GEN Energija we are actively preparing for the potential deployment of SMR technology with a clearly defined roadmap extending toward 2030. In the initial phase, we are focusing on comprehensive technical and economic studies, monitoring developments in reactor designs, and engaging directly with global SMR vendors. Until 2028, the focus is towards a detailed feasibility study and the identification of viable scenarios for the deployment of SMRs in the national energy mix.

Our ambition is clear: by 2030, GEN Energija aims to develop a concrete SMR project concept, suitable for further work and to move rapidly towards implementation when conditions permit.

In conclusion, while Small Modular Reactors represent a promising technology for the future of low-carbon energy systems, their successful deployment will depend on overcoming technical, economic, and regulatory challenges.

During the operation of a nuclear power plant, liquid releases may occur containing both radioactive and non-radioactive substances. In addition liquid releases can represent additional thermal source for the environment. These arise primarily from the cooling systems and systems for the preparation of technical water. A key aspect in the planning of the JEK2 project is therefore the comprehensive management of such releases, ensuring environmental protection, regulatory compliance, and long-term societal acceptability.

Among the radionuclides, tritium (³H) is expected to be the most prevalent, while other fission and activation products will be present in smaller quantities. Non-radioactive substances include minerals, chemical additives, and organic compounds. The anticipated volume of liquid releases containing radionuclides is approximately 250 m³/h, whereas most liquid releases will result from the blowdown of the secondary cooling systems, with an estimated volume of around 2550 m³/h.

All liquid releases will be treated using engineered systems for collection, separation, purification, evaporation, and decay storage, ensuring the reduction of activity prior to discharge into the Sava River. Discharge limits and conditions will be defined in the environmental permit. Provided that these conditions are met and the ALARA principle is observed, the environmental impact of JEK2 on surface water bodies will be negligible.

The JEK2 project is a key strategic investment for strengthening Slovenia’s long-term energy security, reducing carbon emissions, and supporting the transition to a low-carbon energy system. This paper outlines the current project schedule, focusing on the period leading to the Final Investment Decision (FID), planned for 2028. The schedule includes spatial planning, safety documentation, supplier selection, and financing and business model definition.

The timeline covers 2025 to 2041, with commercial operation of the new nuclear unit planned for 2041. Key milestones are the approval of the National Spatial Plan (NSP) and FID in 2028, supplier contract signing in 2029, Integrated Construction Permit and First Concrete Pouring in 2032, followed by operation start in 2041.

The spatial planning process began with the submission of the NSP initiative in October 2024. Formal procedures are expected to start in mid-2025 with the publication in the Spatial Information System (PIS). Public consultations are planned for late 2025, with NSP approval expected in 2028.

Safety documentation preparation started in 2024 with the Site-Specific Safety Analysis Report (SSAR), followed by the radiological impact assessment, Probabilistic Seismic Hazard Analysis (PSHA), and flood safety assessment. These activities are scheduled for completion by the end of 2026 and will support permit applications for construction and operation.

Supplier selection is based on NOAK technology with one reactor unit. The Technical Feasibility Study (TFS), covering key technical risks, will be completed in 2025. Supplier qualification, tendering, bid evaluation, and negotiations are planned to conclude by the end of 2028, aligned with the FID, and contract signing is planned for 2029.

Economic, technical, and macroeconomic analyses have been completed. The next steps include defining financing sources through GEN, GEN Group, Slovenian Sovereign Holding (SDH), the Ministry of Finance, investment banks, technology suppliers, and EXIM banks, while addressing funding gaps and co-investment options.

The schedule is considered feasible, with potential for acceleration through legislative adjustments or special law adoption. Successful project delivery will require strong stakeholder coordination, timely decisions, and effective risk management throughout all phases.

The global shift toward low-carbon and sustainable energy is reshaping how large-scale infrastructure projects are planned, evaluated, and communicated. In this context, nuclear energy is re-emerging as a vital component of national and regional strategies due to its ability to produce stable, low-emission electricity.

The JEK2 in Slovenia faces the challenge of aligning three complementary but distinct sustainability frameworks. The CSRD (Corporate Sustainability Reporting Directive) is a mandatory EU directive that obliges companies to disclose their environmental, social, and governance (ESG) impacts, with a strong emphasis on transparency and accountability. Complementing this, the ESRS (European Sustainability Reporting Standards) define the structure and content of these disclosures, supporting the implementation of the EU Sustainable Finance Taxonomy, which classifies economic activities based on their contribution to environmental sustainability. However, neither CSRD nor ESRS is specifically tailored to the complexities of the nuclear sector. To address this gap, the INPRO methodology, developed by the International Atomic Energy Agency (IAEA), provides a sector-specific framework for assessing the long-term sustainability of nuclear energy systems. Its structure is grounded in the six dimensions defined by the UN Brundtland Commission—including environmental impacts, safety, waste management, infrastructure, economics, and non-proliferation. INPRO enables comprehensive evaluation across the entire nuclear lifecycle and identifies areas where improvements are needed to ensure long-term sustainability.

This integrated approach ensures that sustainability is not merely a reporting obligation, but a strategic foundation for the governance and long-term viability of the project.

The paper will present an in-depth analysis of the long-term availability of uranium in the context of the expanding use of nuclear energy for electricity generation in Slovenia. Based on current data, a comparison will be presented between global uranium resources, annual demand, and future growth projections, with a detailed breakdown of resources according to their reliability, cost-effectiveness, and accessibility. The analysis includes an assessment of identified, estimated, and potential resources and highlights the significant untapped potential for the discovery of new conventional and unconventional sources.

The paper will also provide an overview of current and future trends in the global nuclear fuel market, with a particular focus on the European region. Supply chains, existing conversion and enrichment capacities, and the roles of key suppliers will be analyzed. The diversification of supply sources and the importance of maintaining strategic inventories to enhance supply resilience will be emphasized. The impact of fluctuations in the prices of upstream services (mining, conversion, enrichment, and fuel fabrication) on the overall cost of nuclear fuel and the levelized cost of electricity production will be evaluated, showing that the cost of electricity generation of JEK2 project remains stable despite potential market changes.

Finally, the paper focuses on measures to ensure the long-term security of nuclear fuel supply, including supply diversification, long-term contracting, maintenance of strategic reserves, and participation in international mechanisms such as Euratom. A specific analysis for Slovenia and the JEK2 project will be presented, confirming that the existing studies, available resources, and planned measures are sufficient to ensure energy security, cost efficiency, and a sustainable future.

An important part of the JEK2 project preparation involves analyses needed for planning the management of radioactive waste (RAW) and spent nuclear fuel (SNF), as well as ensuring the safe decommissioning of the plant once its operational lifetime has ended. To ensure continuous, safe, and efficient handling of operational waste, careful planning and a precise assessment of required capacities and associated costs are essential.

The management of low- and intermediate-level radioactive waste (LILW) from the JEK2 project will, from a technological standpoint, represent an extension of the currently planned approach described in the Resolution on the National Program for Radioactive Waste and Spent Fuel Management (ReNPROIG), which serves as the reference in this article. ReNPROIG also includes plan for managing LILW of the Krško Nuclear Power Plant (NEK). Based on preliminary data regarding potential designs for nuclear power plant at the Krško site, a comprehensive management plan for radioactive waste (RAW) and spent nuclear fuel (SNF) has been researched specifically for JEK2. Drawing on findings from various reports and the operational experience, the implementation of the JEK2 project will require an expansion of the existing framework outlined in ReNPROIG. During its operational phase, the JEK2 project will necessitate an increase in disposal capacity, which will involve obtaining the appropriate permits for construction at the current disposal site.

Spent nuclear fuel, classified as high-level radioactive waste (HLW), is initially stored in a spent fuel pool and later transferred to a dry storage facility located on the site of the nuclear power plant. This dry storage will be used to store all HLW and spent fuel until a deep geological repository is established to ensure long-term isolation of the waste from the environment. The technological approach to HLW management for JEK2 will follow the already developed model currently in place. A dry storage facility is planned, and after the decommissioning of JEK2, the HLW will be disposed of in a deep geological repository.

Besides the technical point of view, the article also examines the economic aspect of the funding required for radioactive waste management, based on the established practice of collecting resources through a fund, which serves as a dedicated financial mechanism for managing resources and supporting targeted developmental objectives.

The construction of the second nuclear power plant unit in Krško (JEK2), combined with renewable energy sources, represents a cornerstone in achieving Slovenia's goals for carbon neutrality and decarbonization of the electricity generation sector. JEK2 is envisioned as a modern, reliable, safe, and environmentally friendly energy source that will contribute significantly to supply stability and the affordability of electricity for both large and small consumers.

As Slovenia’s largest infrastructure project, JEK2's economic viability has been subject to early and comprehensive evaluation. GEN energija, the project’s investor, conducted a technical dialogue with potential nuclear technology suppliers during 2022 and 2023 to gather input for the preliminary economic analysis. In May 2024, GEN energija published the results of the Preliminary Pre-Investment Economic Analysis of the JEK2 project. The analysis integrates insights from operational experience at the existing Krško Nuclear Power Plant (NEK), supplier input, conceptual designs for JEK2, and international benchmarks.

The analysis assesses three reactor capacity scenarios—1,000 MWe, 1,250 MWe, and 1,650 MWe—with specific investment costs ranging from EUR 9,310 to 9,587 per kWe. Total capital costs (excluding construction financing) are estimated between EUR 9.6 billion and 15.4 billion. The assumed financing model is a mix of equity and debt, with up to 49% guaranteed by the state. Sensitivity analyses were performed to evaluate the influence of various weighted average cost of capital (WACC) values and debt levels.

Estimated average operational (LCOE) costs for JEK2 range from EUR 41.86 to 45.64 per MWh. The economic analysis, supported by an independent data audit (Ernst & Young) and a peer-reviewed economic model (University of Ljubljana), confirms the project’s feasibility across all power output scenarios. Key influencing factors include electricity market price, capital costs, and WACC, while fuel and labor costs have a lesser impact.

A probabilistic risk analysis complements the sensitivity analysis, identifying key uncertainties such as construction delays, capital cost overruns, and energy price volatility. These findings underscore the importance of robust risk management and justify continued development of the JEK2 project, while highlighting the necessity for detailed risk assessment in future project phases.

This paper provides a comprehensive overview of the financing models and financial mechanisms employed in the development of new nuclear power plants. Given the capital-intensive nature, long construction periods, and high risk profile of nuclear projects, securing appropriate financing is one of the most critical elements for their successful implementation. The article explores a range of financing models, including fully public, fully private, and hybrid structures that blend state support with private investment.

Special attention is given to real-world examples, illustrating how different financing approaches have been applied in practice. These include major European nuclear projects such as Flamanville III in France, Hinkley Point C and Sizewell C in the United Kingdom, Olkiluoto 3 in Finland, Dukovany 5 and 6 in the Czech Republic, as well as current and planned nuclear programs in Sweden and Poland. Each case provides insight into the financial architecture, risk allocation, role of government guarantees, and involvement of energy utilities and private capital.

In addition to the structural models, the article also examines specific financial mechanisms used to support these investments, such as Contracts for Difference (CfD), Regulated Asset Base (RAB) models, export credit agency financing, sovereign loan guarantees, and power purchase agreements (PPA). The paper highlights how the choice of financing and risk-sharing model significantly influences the project's bankability, investor interest, and public acceptance.

By comparing various approaches and mechanisms, the paper aims to inform future decisions on nuclear financing strategies, particularly for countries considering new nuclear builds as part of their long-term energy and decarbonization policies.

Nuclear research reactors (RR) are essential facilities for commercial reactor development and operation, training and education, advancing both fission and fusion technologies, and many other applications not directly related to nuclear energy production (e.g., medical isotope production, neutron sources, material science). The broad and diverse landscape of RRs in Europe is ageing (on average 59 years) and their number is decreasing, due to the recent shutdown of several low- and medium-power reactors. In addition, with the shift towards a zero-carbon society, there has been an increase in demand for RR services, which aligns with Europe’s growing interest in nuclear energy. The VERONICA (Versatile European Reactor fOr Neutron Irradiation and nuclear research) project, a collaboration between “Jožef Stefan” Institute (JSI) and CEA, addresses these challenges by proposing a new versatile European research reactor in Slovenia [1]. The primary objective of commissioning a new research reactor is the development of a comprehensive strategy and an implementation roadmap in line with the IAEA’s “Strategic Planning for Research Reactors” document [2]. The initial tasks for the development of the strategy for implementation that the decision-makers can use, are presented in this paper.

The first task involves reviewing the stakeholder needs by surveying existing European research reactor projects (such as Jules-Horowitz Reactor and PALLAS), collecting data on nuclear energy needs, identifying potential stakeholders and analyzing their specific requirements. As a first step potential stakeholders were identified, both Slovenian and European, each representing a different category, such as Government and Regulatory bodies, Universities and Research Institutions, Industry and private sector, Healthcare and Medical Research, and International Collaborations and Networks. In the second step, utilization was analysed by identifying more than 30 different RR applications in the fields of physics, chemistry, medicine, material and environmental sciences, technological applications, teaching and education, etc. Based on each stakeholder’s needs a connection to each application was identified to create a stakeholder-utilization matrix, highlighting where the demand is concentrated and where strategic partnerships or investments may be needed. The stakeholder-utilization matrix was created for a Zero-power reactor (ZPR) and a Multi-Purpose Research Reactor (MPRR).

The second task, presented in the paper, is the identification and analysis of functional and technical requirements of the new RR. Based on the results from analyzing the needs of different stakeholders, the analysis into the most suitable type of a RR was performed with the emphasis of researching a potential dual-core facility in which both a ZPR and a MPRR could be co-located in the same building, as two perfectly complementary tools. Such a design is in line with the final recommendations for building new European RRs, within the TOURR (Towards Optimized Use of Research Reactors) project [3] and the strategic need for future European nuclear education [4]. By presenting the early insights in framing the broader strategy for VERONICA, this paper aims to contribute to the ongoing European dialogue and invite collaborations and feedback from the wider nuclear science and technology community as the project moves towards its conceptual design.

The presented data will be in the next tasks used for reactor pre-design in which simulations and modelling tools will be used to assess RRs performance and select the most suitable research reactor concept. Last step will be the development of a business model, including technical, economical and societal impact and a potential regulatory analysis in which existing legislation and regulations will be reviewed to develop a plan for potential future Slovenian regulatory requirements mitigation. Such an analysis will position the new RR and its associated services in the European “market” by estimating its added value.

[1] Malec, Jan, et al. "Advancing Nuclear Research and Education in Slovenia and EU: From Operating the TRIGA Reactor to Building a New Generation Facility." Arabian Journal for Science and Engineering (2024): 1-13.

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Strategic Planning for Research Reactors, IAEA Nuclear Energy Series No. NG-T-3.16, IAEA, Vienna (2017)

[3] Pungerčič, Anže, et al. "European research reactor strategy derived in the scope of the towards optimized use of research reactors (TOURR) project." Annals of Nuclear Energy 211 (2025): 110963.

[4] Cizelj, Leon, et al. "Towards strategic agenda for European nuclear education, training, and knowledge management." Nuclear Engineering and Design 420 (2024): 113001.

Large scale energy projects, such as the second unit of the Krško Nuclear Power Plant (JEK2), present serious organisational challenges that go well beyond purely technical and engineering dimensions. To succeed in a referendum on Slovenian nuclear program and start the project of building a new power plant, it is crucial to understand the project as a social phenomenon before addressing its technical aspects. So far, the communication strategies and public engagement approaches employed by energy companies have often proven inadequate, leading to delays in implementation and increased costs. Presentation addresses three key questions that are not only vital for Slovenia’s nuclear program but also for the broader success of the Slovenian energy transition. First question is: How have existing approaches to public engagement performed in the context of nuclear power plant construction? Second question is: How has the public responded to those communication practices? Third question is: What changes are necessary to improve public engagement and the effectiveness of communication? In other words. What must be done to reduce the risks ahead of and after the referendum? These questions are particularly relevant for the discussion in the context of NENE 2025, given the time constraints, as the timeframe needed for scientific and expert analysis is inevitably running out within the project timeline for JEK2.

In this paper, a small modular reactor core design concept using simulated annealing (SA) methodology is studied for optimized loading pattern. Core designers in the field of nuclear engineering invest high manpower and time in searching an optimized core loading pattern that satisfies both safety and economic efficiency. The greatest advantage of using the SA based program is that it can significantly reduce the required manpower and time for searching a core loading pattern. This study intends to show that a SMR core loading pattern candidates selected by SA methodology-based program has sufficient performance for practical use in real core design. To validate the loading pattern derived by the SA based program in terms of core neutronics, it was compared with one of the loading patterns from innovative SMR (i-SMR) currently under development in Korea. The target cycle length of core in this study is more than 2 effective full power years (EFPYs), the SMR core has minimized power peaking factor and reactivity swing.

Making the promise of New Nuclear a reality requires industrial challenge with design industrialization and supply chain development to deliver on time and on budget critical components.
Framatome is ready to support these endeavors.

To avoid uncertainties in managing such important projects, Framatome has standardized components manufacturing and increased its manufacturing capabilities in order to be able to serve the market demand.

- Re-localization of key components’ manufacturing

- Increase of manufacturing capacities to address EPRs and SMRs needs + techno bricks

- Flexibility optimization

With 14 EPRs either in operation or under construction, Framatome addresses the industry challenge going standard. Standardization plus our commitment to deliver every year components for up to 3.5GWe of new nuclear is a powerful lever to de-risk projects and ease financing of new nuclear programs.

With proven technology, full European IP, a qualified supply chain, and available capabilities, Framatome is ready to deliver Large Size Reactors and SMRs nuclear programs.

The high level of nuclear quality and safety requirements impose a trusted expertise of process engineering & manufacturing as well as in-depth knowledge of materials metallurgy quality.

Framatome additive manufacturing solution allows a stable and repeatable quality for manufacturing of critical components, with safer and more efficient supply chain strategy, in order to replace casting and forging conventional solutions, depending on sizes and geometrical complexities.

Relying on its thorough expertise in the nuclear industry, Framatome has developed a know-how in additive manufacturing throughout the whole value chain of components production, including the mandatory demonstration and qualification stages:

- Manufacturing process assessments,

- Expertise of material properties, in-service behaviour studies,

- Optimised design/re-design to additive manufacturing,

- Process expertise & solving technological issues linked to manufacturing quality,

- Manufacturing & quality controls,

- Production & components qualification process management.

Lead-times and costs reductions are powerful levers to re-risk projects and ease financing of new nuclear programs.

With proven technology, full European IP, a qualified supply chain, and available capabilities, Framatome is ready to deliver Large Size Reactors and SMRs nuclear programs.

Small Modular Reactors (SMRs) are gaining interest for grid integration, yet their fuel utilization compared to larger reactors remains understudied.This work compares fuel cycle lengths and conversion ratios (CR) for reactors with thermal powers from 160 to 1107 MW and initial enrichments from 4.95 % to 15 %, using a common PWR design modeled in OpenMC. All cores share a fixed diameter-to-height ratio and use UO2 fuel with added gadolinia in the outer ring. To allow for natural circulation cooling, linear power per fuel rod is conserved across sizes.Fuel depletion is modeled by coupling transport, depletion, and boron adjustment to maintain criticality. Results show larger reactors achieve better fuel utilization, though the gap narrows at higher enrichments. SMRs may require higher enrichment to remain competitive.

Lead- and sodium-cooled advanced small modular reactors (SMR) are being pursued in Generation IV nuclear programs, offering significant advantages over traditional designs. Molten lead and sodium coolants provide high thermal conductivity and high boiling points, enabling efficient heat transfer at near-atmospheric pressure. Both also exhibit favorable neutronic characteristics (e.g., lead’s very low neutron absorption) and have been adopted in planned reactor designs and current research and development worldwide. These systems incorporate passive safety features such as natural circulation cooling and modular construction potential.

Ensuring the structural integrity of these reactors poses unique challenges for in-situ ultrasonic non-destructive evaluation (NDE) under extreme temperatures. Conventional ultrasonic equipment (e.g., PZT transducers) loses functionality beyond ~350 °C or even much less – practically below operating temperatures where these coolants remain molten and functional for the reactor operation (lead >327 °C; sodium >98 °C). Other components (couplants, sensor housings, wiring) likewise risk severe degradation, evaporation, or failure. These constraints necessitate novel approaches to enable reliable high-temperature inspection.

This study presents a detailed phenomenological model (that uses thermodynamics laws and FEM) to predict ultrasonic NDE signal behavior dynamics through reactor vessel walls in contact with molten lead or sodium coolant and as a function of changing temperature. The model characterizes how signal amplitude, wave packet duration, and central frequency vary with vessel wall temperature, effectively capturing coolant-temperature influences on inspection signal quality. The model applies to both lead- and sodium-cooled conditions and highlights distinct thermal regimes, revealing an optimal inspection temperature window. At lower temperatures just above the coolant’s melting point, incomplete melt and high viscosity impair ultrasonic transmission. At higher temperatures, transducer sensitivity and acoustic coupling degrade due to excessive heat. An intermediate temperature range is identified that yields superior signal clarity and amplitude. These insights are essential for planning inspections that ensure structural integrity, maximize signal quality, and avoid equipment damage in advanced reactors using either lead or sodium coolants.

Molten sodium-cooled Small Modular Reactors (SMRs), being developed under the Generation IV nuclear initiative, could offer several advantages over conventional reactor designs. However, ensuring the structural integrity of critical components in these advanced systems will present unique challenges for non-destructive evaluation (NDE) due to their more extreme operating conditions compared to traditional nuclear power plants. In particular, in-service inspection of components such as heat exchanger tubing must contend with molten sodium’s high temperatures and chemically reactive nature. Conventional eddy current testing (ECT) techniques, while standard for inspecting heat exchanger tubes, typically rely on probe materials (e.g., coil wire, insulation) that degrade or fail at temperatures below 150 °C, i.e. below the lower temperature limit of the typical operating range of sodium systems, which can exceed 350 °C. This limitation necessitates the development of novel approaches for reliable ECT performance under high-temperature conditions. This study presents both experimental and modelling investigations of a bobbin-type eddy current probe designed for non-contact, high-temperature operation. The results of this work identify specific temperature thresholds beyond which the probe’s signal-to-noise ratio and flaw detection sensitivity degrade significantly, marking the practical upper limits for current probe materials.

Small Modular Reactors (SMRs) are expected to offer economic, safety and security advantages over current large water reactors. SMRs incorporate advanced safety features, including passive safety systems that rely on natural processes to cool the reactor in accidental situations without the need for active mechanical systems. SMRs can be built in modules, allowing for incremental capacity addition as demand grows, reducing financial risks and construction times. In addition, the use of ATFs in SMRs can allow fuel efficiency to be increased through power uprates and/or higher burnups, deriving benefits both from SMRs and ATFs. Therefore, an I-NERI project involving BNL, UPM, INL and NFQ was proposed and approved with the aim of demonstrating and quantifying the safety benefits of ATF for SMRs. To achieve this, different tools such as neutronic or thermohydraulic models, probabilistic risk assessment (PRA) models and fuel behaviour models are being applied jointly. The aim of this paper is to show the activities planned and the results obtained in the first phase of the project. Planned activities include: development of a TRACE/PARCS model of NuScale; ATF (FeCrAl, Cr-coated Zry, SiC) material properties and thermo-mechanical models; to select and quantify the risk of the sequences of interest and to analyse the impact of incorporating different ATF materials on risk reduction; to assess the safety and performance of different ATF designs in a broad spectrum of anticipated operational occurrences and design basis accidents; to evaluate the performance of ATF cladding and comparing it with the Zry under different transient and accident conditions of interest.

A boron-free small modular reactor (SMR) core has been developed using LEU+ fuel—enriched up to 10 w/o ²³⁵U—in combination with integral burnable absorbers composed of UO₂-Mo-Gd₂O₃, with the objective of reducing the total number of spent fuel assemblies. The adoption of LEU+ fuel, which has a higher enrichment than the 5 w/o ²³⁵U used in conventional PWRs, enables an extended core cycle length and higher fuel discharge burnup. The core design incorporates UO₂ fuel with up high content gadolinia (Gd₂O₃) as a burnable absorber. To mitigate the associated degradation in thermal conductivity due to gadolinia, molybdenum (Mo) is added, thereby improving thermal performance.

To enhance fuel utilization, a partially three-batch reloading strategy is employed, in which thrice-burned fuel assemblies are positioned in the peripheral region of the core. The resulting core achieves a cycle length of 30 months while maintaining key operational parameters—such as the peak power factor and axial offset—within the design limits established for SMR cores utilizing fuel enriched below 5 w/o ²³⁵U. These results demonstrate that the application of LEU+ fuel with UO₂-Mo-Gd₂O₃ burnable absorbers is a viable and effective approach for boron-free SMR designs, offering high burnup capability, long-cycle operation, and improved fuel efficiency.

Micro Modular Reactors (MMR) are nuclear power plants with max 10MWe for applications in remote areas without connections to the grid. A case study carried out by LANL focusses on a solid stat core, liquid metal heat pipes to transport the heat to power conversion unit. In the MISHA project, funded by BMBF, the feasibility of this reactor will be assesses and some basic safety analyses will be carried out. As part of the MISHA project, several potassium heat pipes have been designed and fabricated. The first prototype tested features an internally grooved structure to create capillary forces. Experimental investigations are conducted in a test rig, allowing control over inclination angles and power input distributions. The heat pipe was subjected to a range of operating conditions to evaluate thermal performance.

Notably, geysering phenomena are observed under specific configurations, indicating complex two-phase flow dynamics. Detailed analysis are carried out to understand the onset and characteristics of these instabilities. The experimental findings provide insight into the impact of geometry and operating parameters on heat pipe behaviour. Comparative assessments with existing literature data are performed to validate results and identify deviations.

The study contributes to the creation of reliable experimental data for development and validation of the thermal hydraulics code ATHLET in the framework of the MISHA project.

This work presents a digital twin framework to study how noble metal fission products move and build up in molten salt reactors (MSRs). These elements, formed during fission, tend to stick to metal surfaces or travel with bubbles in the salt, and they can release a noticeable amount of decay heat after the reactor is shut down or drained. Understanding where they go and how much heat they produce is important both for safety and for designing tests in future reactors.

The framework links together three main pieces: neutron transport, depletion, and material flow from Serpent 2; system-level thermal–hydraulics from Modelica; and a Python interface that keeps the two codes in sync. Deposition on surfaces and bubbles is handled with simple 0-dimensional correlations, which are quick to compute but still flexible enough to cover a wide range of conditions. This makes it possible to study reactor-scale effects without the heavy cost of full CFD.

To check the method, the results were compared with data and experience from the Molten Salt Reactor Experiment (MSRE). Even though MSRE used different fuels (U-235 and U-233) and its records have uncertainties, the framework shows consistent deposition patterns and decay heat behavior. This gives confidence that the approach is reliable enough to support design work.

The framework was then applied to a commercial MSR case, looking at what happens after draining in the primary heat exchanger, a component expected to be most affected by deposits. In this situation, decay heat from noble metals can reach a few hundred kilowatts. To deal with this, a test procedure is proposed that combines zero-power periods and low-power operation with sinusoidal power changes. The results show that this method can separate the heat coming from deposits without requiring a full shutdown.

Molten salt chemistry of tellurium is of increasing interest for applications in advanced nuclear systems, particularly molten salt reactors (MSRs), where tellurium’s reactivity and potential volatility raise concerns for material corrosion and fission product management. This study presents experimental data on the behavior of tellurium species in a molten NaCl–MgCl₂ matrix at 500°. Using electrochemical analysis, we investigated the redox behavior and stability of metallic Te, commercial Na₂Te, and Na₂Te synthesized in situ via reduction of Te with aluminum. Cyclic voltammetry and open circuit potential measurements revealed multiple redox systems, including Te⁰/Te²⁻ and Te⁰/TeCl₂ transitions, and highlighted the dynamic equilibrium between dissolved and deposited tellurium species. Long-term experiments showed that telluride species gradually disappear, likely due to volatilization. ICP analysis confirmed the presence of tellurium deposits on colder system surfaces and XRD analysis revealed its metallic speciation. These results offer new insights into the complex redox and transport behavior of tellurium in chloride melts, supporting the development of more robust MSR designs.

Framatome Inc. Presentation for the NENE 2025 Conference

September 8^th-11^th 2025, Slovenia

To the attention of the Organization

Framatome Reactor Service Tooling Innovation

As nuclear power plants continue to age so does the Outage tooling utilized to service them. With advancements in technology, Framatome has developed new innovated tooling to replace legacy OEM tooling which has become inefficient and costly to refurbish. The tooling focuses on increased production, elimination of historical FME issues, and overall simplification for the end-user.

This PowerPoint presentation explores several case studies in which new industry tooling has been developed to replace antiquated legacy industry tooling. Additionally, the presentation will discuss how to develop a simple business case which shows the return on investment including cost savings to help utilities receive funding internally. Finally, the presentation will discuss several outside of the box approaches to perform unique examinations in which innovation was included to perform multiple activities while on critical path to reduce outage schedule time.

The overall goal of the PowerPoint presentation is to show how innovation doesn’t need to be complex. Often innovation can be focused on simplifying legacy tooling to provide a more user friendly product leading to increased production.

Presenter: Kris Cecil

COMPLEX EVALUATION OF THE LIFETIME AND CHEMICAL REGIME OF STEAM GENERATORS

Ing. Jerhotová K., Mgr. Joanidisová M., Ing. Kučerová T.

ÚJV Řež, a. s., Hlavní 130, Husinec-Řež, Czech Republic, katerina.jerhotova@ujv.cz

With the increasing demands for safe and long-term operation of nuclear units, the requirements are especially aimed at increasing the quality of used materials, improving the monitoring of physical-chemical processes, modernization of technological systems and making the service operation more effective. By reducing the repair and maintenance time, it is possible to reduce costs and the potential radiation exposure to workers.

Evaluation of the chemical regime of secondary circuit as well as a comprehensive evaluation of its history is essential for its understanding and control. In order to set a proper chemical regime and optimal operating conditions in the secondary circuit, especially in steam generators (SG), various methods can be used:

• Monitoring and measuring chemical parameters in feedwater and blowdown during the campaign,
• NDT inspections of heat exchange tubes in SG to prevent primary to secondary leaks,
• Scraping samples from the tube surfaces - to determine the composition of surface layer on heat exchange tubes via SEM (EDS method),
• Chemical analysis of samples from the SG stand (SG stand is a device, that demonstrates the crevice environment inside the SG, and it can be connected to the main feed water distribution pipeline of the secondary circuit. A liquid samples can be obtained, analysed and the high temperature pH(t) in crevices can be calculated),
• Hide out return calculation - calculation of high temperature pH(t) in crevice environment during the campaign based on the samples obtained during the shutdown of the unit.

During unit shutdown, high temperature crevice pH(t) can be determined retrospectively from blowdown samples. An important parameter for the release of impurities is the temperature of the primary circuit at which the reactor reaches low power levels. With the decreasing temperature, the boiling point in SG begins to disappear, water penetrates into the crevices and dilution of impurities follows. As a result, an increased concentration of impurities in blowdown can be observed (Hide out return) and a high temperature pH(t) can be calculated. The basic and the most important species to be monitored and measured are Na⁺, K⁺, Ca ²⁺, Mg²⁺, Cl^-, SO₄ ^2-, SiO₂. Optional species for evaluation are NO₃^-, F^-, Fe, TOC.

The high-temperature pH(t) depends on the concentration of the solution when the vapour is formed. It has been proven that a neutral high temperature pH(t) is optimal for the Steam generator operation. Concentrated solutions inside crevices can be aggressive from a corrosion point of view and can damage the SG tubes material (tube surfaces for the secondary side). The resulting pH(t) of the solution depends on the balance of cations and anions in the SG water.

High amounts of calcium, magnesium, sulphate and silicate ions can lead to formation of precipitates. Solids contribute only partially to the character of the crevice pH(t). Their presence supports the formation of partially closed areas, and thus promotes the concentration of other salts (impurities) in the SG crevices.

I. INTRODUCTION

The need for flexible operation of nuclear power plants is becoming critical. It is driven by the expanding role of nuclear power as a large-scale, economical, carbon-free energy source, as well as the growing integration of intermittent renewable energy sources such as solar and wind power into the electricity grid. To facilitate such operational flexibility, an online monitoring system that can assess the real-time condition of the reactor core is essential.

In response to these demands, KEPCO NF has been developing a high-performance monitoring system based on its proprietary code package. One of the key features implemented in the system is the 3D Power Connection Method (PCM). It synthesizes the 3D intra-core power distribution using a limited number of signals from fixed in-core Self-Powered Neutron Detectors (SPNDs).

Well-established methods such as CECOR and INCORE have been used to synthesize the 3D power distribution of large commercial PWRs. On the other hand, the 3D PCM employs 3D Power Connection Factors (PCFs) that are continuously updated during reactor operation. This updating scheme allows the method to accurately capture the real-time core state.

In this abstract, the enhanced 3D PCM, which improves robustness of the 3D PCM, is briefly introduced and the calculation results for a xenon oscillation scenario on a simulated OPR-1000 reactor core are presented. More details and further scenarios will be included in the full paper.

II. ENHANCED 3D POWER CONNECTION METHOD

The 3D nodal power distribution can be obtained by using a nodal diffusion code. Based on the distribution, the PCF at each node (i,j,k) is defined as shown in Eq.1. Once calculated, the PCFs are used until the next update.

PCF_(i,j,k)=(∑▒P_(i,j,k)^n )/(P_(i,j,k) N_(i,j,k)^nb ),n=B,N,W,E,S,T
(1)

PCF_(i,j,k) and P_(i,j,k) denote the PCF and power of the target node, while N_(i.j.k)^nb represent the number of neighboring nodes. The superscript B, N, W, E, S and T denote the values of nodes neighboring the target node at the bottom, north, west, east, south and top directions, respectively. Eq.2 is derived for each of the N nodes by employing the PCFs, resulting in a system of equations represented by Ax=b, where A is an N×N matrix.

P_(i,j,k)^B+P_(i,j,k)^N+P_(i,j,k)^W+N_(i,j,k)^nb PCF_(i,j,k) P_(i,j,k)+P_(i,j,k)^E+P_(i,j,k)^S+P_(i,j,k)^T=0 (2)

In addition, each measured power is the volume-weighted sum of the power at the corresponding SPND-instrumented nodes. Therefore, Eq.3 is satisfied. Note that dk, P_dk, and V_dk denote the ID, measured power, and volume of the SPND.

∑_(i,j,k∈dk)▒〖V_(i,j,k)/V_dk P_(i,j,k) 〗=P_dk (3)

Under these constraints, the Lagrange multiplier method is applicable. This is one of the key features of the enhanced 3D PCM. Previously, the assumption that SPND-instrumented node powers were known caused the synthesized power error to grow significantly over time from the PCF update point.

However, it should be noted that the PCFs are derived from the code calculation and represent only engineering estimates of the actual core state. Therefore, directly applying the Ax=b system may introduce unphysical fluctuations in the synthesized power due to inconsistency between the calculated results and the SPND measurements. To address this, the objective function within the Lagrange formulation is modified to use A^T A instead of A. This modification improves the solution by reducing the impact of the inconsistency.

For verification of the enhanced 3D PCM, a free xenon oscillation simulation was performed at the end of cycle (EOC) for a modeled first-core of the OPR-1000. The oscillation was initiated by inserting and subsequently withdrawing the control bank. The simulation covered 32 hours with a time step size of 30 minutes. The Advanced Static and Transient Reactor Analyzer (ASTRA) code was used to generate time step-wise reference 3D power distributions and corresponding SPND signals. The PCFs were obtained from the core at equilibrium prior to the onset of oscillation and were intentionally not updated throughout the simulation. It was intended to assess the capability of the enhanced method.

For the root-mean-square errors of the synthesized nodewise 3D power (P3D) and core-averaged 1D power (P1D), and the peaking factor errors (ΔF_q, ΔF_xy), the errors remained below 1% for the majority of the simulation. It must be emphasized that, in the actual core, the PCFs are updated every 10 seconds even during transient states keeping the error close to 0%. The presented case intentionally assumes an extreme and unrealistic scenario, only to demonstrate the robustness of the enhanced 3D PCM.

III. SUMMARY

While the 3D PCM provides high accuracy and computational efficiency, its applicability has been constrained by potential inconsistencies between the calculated PCFs and the actual core condition. This study proposes an enhanced approach to improve robustness as well as to extend temporal coverage. The full paper will present verification results under various transient conditions, including control rod drop and steam generator trip scenarios.

This paper explores solutions for modernizing Instrumentation and Control (I&C) in nuclear and military settings, with a focus on digitization. The goal is to replace outdated analog instruments and systems with digital technologies in a way that will minimize operational interruptions. The challenges associated with digitizing these complex systems can be addressed through two key approaches: the use of universal adapter plates for gradual meter replacement and the implementation of a modular compact I&C cabinet for comprehensive overhaul and both options compliant to Class 1E or commercial grade.

The adapter plate solution facilitates the transition from analog to digital meters by converting existing panel cutouts to accommodate a range of digital displays. This approach allows facilities to replace outdated meters without the need for a complete new control panel. These adapter plates offer the flexibility to perform upgrades incrementally, either as analog meters fail or during planned maintenance outages. The digital meters provide improved accuracy, reliability, longevity, and compatibility with modern Human-Machine Interface (HMI) and Machine-Machine Interface (MMI) systems. Additionally, converting all meter input signals to a standardized 4-20mA current loop simplifies maintenance by reducing the variety of costly spare parts required, enabling the use of a single meter type for various signals across the facility.

A more comprehensive solution is the implementation of a modular compact I&C cabinet, which centralizes the control of multiple I&C functions into a single, unified system. Such a cabinet allows for full or partial digitization of control systems without interfering with ongoing plant operations. This modular system can be installed in any available space with access to power and signal connections, and can function alongside or as a replacement for existing Supervisory Control and Data Acquisition (SCADA), Data Acquisition Systems (DAS), or Programmable Logic Controllers (PLC). A digital I&C cabinet supports both automated and manual control of process, enhancing flexibility in operation. A digital I&C cabinet would be required to be designed to house a large number of digital meters, controllers, and signal transmitters. Desired features include programmable logic transmitters and controllers that can manage signal outputs via various relay options including an OCT or MOSFETS or dry contact reel relay up to 4 amps for Hi, Hi/Hi, Low, and Low/Low alarms, and fail-safe mechanisms. These systems should be fully programmable and customizable to meet specific operation needs. The development and deployment of modular I&C cabinets have been informed by lessons learned from various industries, including nuclear power, aerospace, and military applications. These robust steel cabinets construction provides housing for up 400 digital meters, while redundant UPS and individual isolated power supplies and advanced monitoring systems ensure operation reliability. Sophisticated fail-safe capabilities are crucial for critical operations and must include mechanisms for controlled shutdowns during emergencies. This approach to digitizing I&C systems can be customized to meet the specific needs of both civilian and military applications. The system’s ability to consolidate control functions into a centralized digital cabinet eliminates the need for multiple scattered control panels, streamlining operations and improving overall system performance. Furthermore, the use of standard current loop signals simplifies maintenance and reduces long-term operational costs by extending the lifespan of the system and its components.

The lack of a domestic research reactor in the United Kingdom (UK) has profound consequences for its ability to source radionuclides for medical applications, relying on the European mainland to satisfy its medical demands. Reactors such as BR2 (SCK, Belgium) and the HFR (Petten, Netherlands) have historically been adequate to meet healthcare needs. However, these reactors are reaching the end of their life, with BR2 scheduled to be decommissioned no later than 2036. Other reactors are likewise reaching the end of their life, which threatens the continuity of large-scale production for critical radionuclides such as Tc-99m and I-131, essential for medical imaging as well as brachytherapy. As global demand increases and the supply chain is expected to contract, scarcity is anticipated for less common radionuclides, potentially putting patients at risk. To counter this problem, the UK is exploring several pathways to increase its domestic production capabilities. Proton accelerators offer a variety of benefits, including the ability to produce many different radionuclides without the high capital investment associated with high-flux reactors or the large inventory of nuclear fuel needed to run them. Protons, as charged particles, are much easier to shield, and the tendency of (p, xn) reactions to dominate greatly reduces nuclear proliferation concerns. The University of Bristol has been working since 2022 on novel proton accelerator technology using nuclear fusion reactions to boost proton energy. This approach potentially enables more compact designs that cost less to operate and maintain, with a substantial decrease in capital investment requirements. Such smaller accelerators could thus be scaled to the needs of individual medical institutions, bringing production closer to the patient. This work presents the progress made on the prototype, an update on the safety case and regulatory situation, and an outlook for future work.

The divertor tokamak test (DTT) facility is under construction in Frascati, Italy with the objective to support ITER and future fusion power plants on the power exhaust challenges. As a tokamak machine, DTT will include the first wall (FW) as well as the divertor as plasma facing components inside the vacuum vessel. While the operational loads of the FW include a substantial amount of heat from plasma radiation and particle bombardment, the electromagnetic (EM) loads due to disruptive events are anticipated to be a major contributor to the FW integrity.

In this paper, the structural analyses of two supports of the DTT outboard FW are performed under the EM loads that originate from a fast Vertical Displacement Event (VDE) of the plasma. The effective forces and moments due to EM loads on major components of the FW (such as the plasma facing units and support plates) are applied to the finite-element models of the studied supports as remote loads. Additionally, the EM loads on the supports themselves are considered as body forces. The results of the analyses performed with the ABAQUS code include the displacements and stresses acting on the supports, with the main aim to check their structural integrity and to finalize their engineering design.

Accurate calculation of the shutdown dose rate (SDDR) is of crucial importance for evaluating of radiation safety in fusion reactors. This study explores the application of the Rigorous Two-Step (R2S) method and the Direct One-Step method (D1S) determining the SDDR by comparing two computational codes that implement these two approaches.

The JSIR2S code, developed at the Jožef Stefan Institute integrates the particle transport simulations of the MCNP code with isotopic activation and transmutation calculations performed with the FISPACT inventory code. OpenMC, an open-source neutron transport code developed at the Massachusetts Institute of Technology (MIT), has implemented the D1S capabilities for SDDR calculations. The primaryy objective of this study is to validate the JSIR2S code for fusion reactor applications and to gain practical experience with an open-source workflow for SDDR analysis.

A simplified tokamak model in MCNP was used for the calculations and converted for use with OpenMC. The geometrical and material definitions were carefully adjusted to maintain consistency between the neutron transport simulations. Both models were exposed to a deuterium-deuterium (DD) plasma source.

In this study, the shutdown dose rate of the simplified tokamak model was computed with OpenMC and compared against results obtained with JSIR2S. Calculations were performed across various cooling times identify the appropriate time frame for reactor maintenance. The insights gained from this comparative analysis enhance our understanding for the further advancement of safety assessments for future large-scale fusion reactors.

Water as a primary coolant will play an important role in the performance of fusion reactors, as after being irradiated and activated, it causes an ionising radiation field throughout the facility. Therefore, additional protection and shielding for the instruments and personnel must be adequately considered. In direct support of ITER, a new irradiation facility called KATANA was successfully constructed and commissioned in 2024 at the TRIGA Mark II research reactor at the Jožef Stefan Institute in Slovenia. KATANA utilises a closed-water activation loop, which allows to perform various experiments based on water activation and serves as a well-defined and stable high-energy (6 MeV – 7 MeV) gamma and lower intensity (1 MeV) neutron source. The ultimate goal of KATANA is to perform benchmark-quality experiments for validating fluid activation codes and establishing itself as a reference facility for high-energy gamma-ray detector calibration, supporting ITER and future fusion reactors.

As part of the commissioning phase, a series of first experiments were carried out to assess the operational characteristics of KATANA that will serve as the basis for later, more detailed (benchmark) experiments. In this work the response to a change in reactor power and the response to a change in flow rate are presented. Measurements of several quantities based on gamma spectrometry, i.e. counts-per-second (CPS) for a 6 MeV and 7 MeV gamma peak of ¹⁶N with a suitable region-of-interest and using different detectors, i.e. two HPGe and two LaBr detectors, were investigated simultaneously. As expected, a linear correlation was observed between reactor power and gamma emissions, with HPGe detectors showing higher efficiency than LaBr detectors. However, the measured branching ratio of the ¹⁶N decay deviated by about 20 % from the theoretical values, which is attributed to the lack of an absolute detector calibration. This emphasises the need for further detailed calibration experiments.

The effect of the water flow rate on the detector signals was also analysed and a non-linear relationship was found. At lower flow rates, the gamma emissions of ¹⁶N increased steeply, with a saturation point near 0.4 L/s, consistent with the simplified model predictions. The continuous and smooth behaviour is observed as expected, with a notable exception at 7 MeV. In addition, the expected slight decrease in CPS at the higher flow rates has not yet been observed, as seen in the pre-analysis results. Although the statistical uncertainty was below 1 % for the 6 MeV peak and below 6 % for the 7 MeV peak, the total experimental uncertainty still needs to be addressed in future work. These results suggest that working at or above the saturation point is optimal to minimise the uncertainties caused by flow fluctuations.

The KATANA facility demonstrated the desired operating characteristics in the form of high and stable water flow rates, which led to high activity values of the observed activated isotopes ¹⁶N, ¹⁷N and ¹⁹O, which is essential for minimising the experimental uncertainties. First experiments performed at KATANA have provided in-depth knowledge and capabilities for operation and, most importantly, have provided important data to serve as a basis for further, more detailed and advanced experiments.

High-energy γ radiation fields generated by the KATANA water activation facility at the Jozef Stefan Institute present significant dosimetry challenges, as accurately quantifying doses from activated water is difficult. This problem becomes even more critical for future water-cooled fusion facilities, where neutron-induced activation will be orders of magnitude higher than in fission systems. Furthermore, these outcomes could benefit facilities operating with high-energy, high-fluence γ fields. In this work, Monte Carlo–based response calculations were performed for RADFET dosimeters in 6 MeV and 7 MeV γ spectra within the activated water loop of the JSI TRIGA reactor.

A detailed reactor model—including the TRIGA core, the primary irradiation loop (“outer Snail No. 1”) and the KATANA auxiliary pump—was employed to simulate dose rate and energy deposition at discrete locations for RADFET sensors both unshielded and enclosed in lead. Unshielded and Pb-shielded configurations were compared to determine how shielding alters dose measurements, and an activated-pump source term was incorporated to investigate its influence on the calculated dose distributions in every configuration.

Resulting response matrices have been formulated in such a way that they can be directly integrated into RADFET calibration protocols for monitoring high-energy radiation fields. This methodological framework has been developed to facilitate adaptation to varied shielding materials, geometries, and energy spectra. A blueprint has been established for experimental validation at the KATANA water activation facility and for transfer to future fusion research facilities, where precise measurement of activated water dose rates will be critical for safety assurance, diagnostic operations, and component qualification.

Fusion reactors – tokamaks – are complex machines usually surrounded with numerous components and experimental equipment, forming a very complex environment. A typical tokamak exhibits inhomogeneous geometry with larger number of ducts and streaming paths through which neutrons can leak from the torus. It is very challenging to calculate the neutron field outside of the vacuum vessel.

The work on simulations of the neutron field at larger distances from the plasma on a model of a typical large tokamak is presented. In particular, the fluence of neutrons passing through the penetrations of the tokamak vacuum vessel was calculated in order to assess the capability of state-of-the-art numerical tools to correctly predict the radiation streaming in large and complex geometries. The calculations are performed in a two-step process using the deterministic code ADVANTG to determine the variance reduction parameters and with MCNP for subsequent calculation of the neutron field with the Monte Carlo method. The main streaming paths are identified using the study of individual particle tracks, available in MCNP.

It was shown experimentally for tungsten, that He retained close to the surface influences the transport and retention of hydrogen isotopes (HI). Namely, recent experiments studying the interaction of HIs and He, using He seeded D plasmas, showed that the addition of He leads to reduced blistering accompanied by a reduced D retention [1].

Recently, we have performed a systematic series of D exposures for tungsten where He was pre-implanted near the surface. 20 MeV W irradiation was performed before or after the He implantation to create defects within the first 3 µm. The defects created by W ions were used to trap penetrating D through the He surface layer and make it hence possible to quantify D transport beyond the He layer using 3He nuclear reaction analysis. Deuterium and helium depth profiling showed increased D retention at the depth where He was implanted. Nevertheless, retention in the bulk was reduced five times as compared to a He-free reference sample [2]. From this we could conclude that the observed D retention in He irradiated samples is due to D trapping at He bubbles and the microscopy structure promotes D diffusion back to the surface which enhances D re-emission. The same methodology of He irradiation and D exposure was applied to EUROFER97 to clarify the effect of surface-near helium on deuterium transport into and retention in the bulk. To quantify the influence on D uptake at the surface, He was implanted into EUROFER97 samples close to the surface with 1 keV ions with different fluences and at different temperatures. Samples were then exposed to a low flux, low energy (100 eV/D) D ion beam at 370 K. One He-free W-irradiated EUROFER97 reference sample was also exposed to low energy D ions for comparison. The defects created by W ions trap penetrating D and make it hence possible to quantify D transport into the layer below the He layer using ³He nuclear reaction analysis (NRA). Measured D depth profiles show reduced uptake of D in the He-irradiated samples with major D retention near the surface where He is expected to be implanted. The reduction of D uptake is reduced by a factor of two compared to He-free sample. Surface analysis of the possible He bubbles are ongoing for the sample with the largest He fluence of 5.5 ×10²¹ He/m². Results will be discussed and compared to tungsten.

[1] M. Baldwin et al. Nucl. Fusion, vol. 51, p. 103021, 2011 and Nucl. Fusion vol. 57, p. 076031, 2017

[2] Markelj et al. in preparation

The long-term viability of magnetic-confinement fusion hinges on how the plasma interacts with the materials that line the vessel. In a tokamak, imperfect confinement allows energetic particles to drift from the core into the scrape-off layer (SOL), where they intersect plasma-facing components (PFCs). In the SOL, cross-field transport is dominated by intermittent, filament-like structures—“blobs’’—that detach from the confined edge, propagate radially outward, and connect along the field lines to the divertor targets. Each filament delivers a short, localized burst of particles and heat, producing two principal effects: (i) thermal loading that accelerates material fatigue and (ii) physical sputtering that releases wall atoms into the plasma [1, 2]. For tungsten PFCs, the latter is particularly critical; even trace amounts of high-Z tungsten entering the core can radiatively cool and dilute the fusion plasma.

Accurate erosion predictions therefore require knowledge of the energy and angular spectra of the species striking the wall—fuel ions as well as heavier impurities. In the SOL these spectra deviate markedly from Maxwellian because the mean free path rivals or exceeds machine dimensions and collisions are often non-elastic. Fluid or “steady-state’’ assumptions thus misrepresent the actual fluxes produced by blob dynamics.

To address this, we have extended the fully kinetic 1d3v code BIT1 [3] to model time-dependent, filament-driven source. The code resolves the self-consistent evolution of particle trajectories, velocity distribution functions, and sheath interactions along a magnetic field tube, while incorporating realistic sputtering yields that depend on instantaneous impact energy and angle. We generated a series of synthetic blobs spanning experimentally relevant widths and peak temperatures, allowing us to probe how filament scale influences tungsten erosion.

The simulations key finding reveals to be that the transient nature of blobs amplifies the high-energy tail of the impact spectrum compared with an equivalent time-averaged source, raising instantaneous sputtering yields by up to an order of magnitude.

Taken together, these results demonstrate that kinetic, time-resolved treatment of filamentary transport is essential for reliable lifetime estimates of tungsten PFCs and for setting tolerable impurity limits in next-step devices such as ITER and DEMO.

[1] V. Philipps, “Tungsten as material for plasma-facing components in fusion devices”, J. Nucl. Materials 415 (2011), S2-S9, https://doi.org/10.1016/j.jnucmat.2011.01.110.

[2] A. Huber, S. Brezinsek, V. Huber et al., “Erosion and screening of tungsten during inter/intra-ELM periods in the JET-ILW divertor”, Nuclear Materials and Energy 25 (2020), 100859, https://doi.org/10.1016/j.nme.2020.100859.

[3] Tskhakaya, D. "Implementation of dressed cross-section model into the BIT1 code", Eur. Phys. J. D 77, (2023), 135, https://doi.org/10.1140/epjd/s10053-023-00682-w.

We present a robust, automated multiphysics simulation framework developed for high-fidelity modeling of small modular reactor (SMR) cores, focusing on the NuScale design. The simulator integrates Monte Carlo neutronics (Serpent 2) with thermal-hydraulic feedback and depletion capabilities, enabling detailed analysis of core behavior under various operational states including Hot Zero Power (HZP), Hot Full Power (HFP), and Burnup (BRN). Simulation logic is controlled via external batch scripts and a structured input file defining core setup, fuel assembly library and core state characteristics. Key multi-physics coupling is achieved through the use of Serpent multi-physics interface (IFC). The framework supports axial segmentation of each fuel assembly, enabling power, temperature, and material feedback calculations in each axial region independently. Its modular design supports expansion to reduced-order modeling, multi-cycle core follow, and transient analysis.

This study introduces an efficient two-step calculation procedure for PWR core design by integrating Monte Carlo and pin-wise diffusion methods. The methodology combines Monte Carlo's high-fidelity cross-section generation with pin-wise diffusion's computational efficiency to model neutron flux and power distribution in reactor cores. The approach incorporates Super Homogenization (SPH) factors to enhance neutron flux heterogeneity modeling, addressing the complexities of modern reactor designs with advanced burnable absorbers and control rod strategies.

Verification using the APR1400 benchmark demonstrates accuracy comparable to whole-core transport codes while maintaining computational efficiency. The methodology is also applied to innovative Small Modular Reactors (i-SMR), particularly evaluating cores with advanced fuel management and soluble boron-free operations. Results show accurate predictions of neutron flux and power distributions in i-SMR cores incorporating advanced burnable absorbers like HIGA (Highly Intensive and Discrete Gadolinium/Alumina Burnable Absorber). The approach effectively addresses i-SMR-specific challenges, including maintaining reactor criticality during extended operational periods. Through optimized parallelization, 3D reactor calculations are completed within seconds, ensuring practical applicability in various operational scenarios.

This methodology represents a significant advancement in reactor core analysis, offering a high-precision, computationally efficient solution for modern PWR and i-SMR core designs, while maintaining exceptional accuracy in predicting core physics parameters.

The structural materials of a nuclear reactor become radioactive during its operation through a process called activation, which is the absorption of neutrons by stable nuclei. These activated materials represent a significant portion of radioactive waste, and their quantity must be determined during decommissioning. Recently, reactor development has shifted toward small modular reactors (SMRs), which are factory-built and designed for rapid on-site installation. However, their smaller size results in a less efficient neutron economy, meaning a higher fraction of neutrons escapes the core per unit of power. Consequently, more interactions occur with the surrounding structural components.

This work presents calculations of the activation of core components in the SMR NuScale Power Module. The calculations were performed using the Monte Carlo transport code MCNP and the JSIR2S code, which couples MCNP with the isotopic inventory calculation code FISPACT-II. The resulting activity and neutron leakage values for the selected SMR are compared with those of a larger reactor at the Krško Nuclear Power Plant.

For a reliable assessment of the safety and efficiency of nuclear reactors, it is essential to properly account for uncertainties in nuclear data. In this study, these uncertainties are propagated through Monte Carlo neutron transport and depletion calculations using the SERPENT2 code. The SANDY code package is employed to generate 500 perturbed samples of nuclear cross sections, decay data, and both independent and cumulative fission yields, based on the ENDF/B-VII.1, ENDF/B-VIII.1, and JEFF-3.3 nuclear data libraries. The convergence and consistency of the sampling process are validated using the LEU-COMP-THERM-006-001 benchmark.

Covariance data, extracted from the nuclear data libraries, are integrated with the SANDY code to produce the perturbed datasets. These sampled datasets are then used in two independent low-statistics, 2D pin-level Monte Carlo simulations of Krško NPP fuel in SERPENT2, each with a different random seed. This methodology generates correlated sets of key observables — the effective multiplication factor (keff), nuclide inventory, and decay heat — without requiring a large number of computationally expensive high-statistics calculations. Uncertainty estimates are subsequently derived from the correlation coefficients between the two calculation sets, following the so-called GRS method. Finally, the nuclide vector from a reference pin level calculation is used with the same approach to estimate uncertainty of keff for a generalized disposal canister.

The concept of the Dispatchable Adaptive Reactor With Interchangeable compoNents (DARWIN) is aimed at the flexible needs of the modern world in terms of the survival of most adaptable species. The aim is to develop an adaptable and versatile reactor design rather than optimising it for a single purpose. Such a reactor could fulfil specific needs and provide solutions for tasks such as flood pumping, district heating, medical isotope production or desalination.

The configuration of fuel rods and fuel assemblies is hexagonal and light water is assumed as coolant. The calculations are performed in two dimensions with the Monte Carlo transport code Serpent 2, using only fuel assemblies with only one type of fuel without burnable poisons and without considering the entire reactor vessel.

An analysis was performed with different types of nuclear fuel suitable for light water-cooled reactors. Two main categories of fuel were considered. The first includes ceramic fuels such as uranium dioxide (UO₂), mixed oxide fuel (MOX), uranium- zirconium hydride (UZrH_X) and uranium yttrium hydride (UYH_X), each evaluated at different enrichment levels and mixing ratios. The second category comprises metallic fuels, in particular the isotopes ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu.

To evaluate the thermal feedback effects, a parametric study was conducted in which the temperatures of the fuel and coolant were varied independently. This approach allowed a detailed investigation of the individual contributions to the overall reactivity feedback, which is crucial for reactor safety and performance optimisation.

This study examines zirconium hydride (ZrHx), a material widely used in research reactors, including TRIGA systems. The hydrogen bound in its crystal lattice plays a crucial role in the behaviour of hydride fuelled reactors, so accurate modelling of reactor performance is highly dependent on the quality of nuclear data related to thermal neutron scattering. In addition, zirconium hydride has attracted interest due to its potential applications in innovative fuel concepts for next generation reactors. This interest is underscored by the fact that thermal scattering law (TSL) data in existing libraries is outdated and benchmark results for fuel assemblies with zirconium hydride show considerable scattering.

Zirconium hydride (ZrHx) exists in different phases, with the δ phase (1.56 < x < 1.64) and ε phase (x > 1.74) being the most important at room temperature. In this study, density functional theory (DFT) incorporated into the VASP code was used to derive force constants, used in the Phonopy and NJOY codes for lattice dynamics and thermal scattering analysis. In contrast to ENDF/B-VIII.1, older data libraries do not distinguish between the ZrHx phases. The main results of the present work include detailed insights into atomic structures, mechanical properties, phonon density of states, thermal scattering cross sections and benchmark results, which are compared with data sets (e.g. ENDF/B-VII.1, JEFF-3.3, ENDF/B-VIII.0 and ENDF/B-VIII.1), experimental data from various literature sources and with other calculations from the literature.

The phonon density (pDOS) reflects the vibrational modes of materials across different energy levels. For ZrHx, the hydrogen spectrum falls predominantly in the optical region, with a smaller proportion in the acoustic region, while zirconium shows the opposite pattern. Analysis of the thermal scattering cross sections shows that the elastic cross sections agree in all evaluations, except for zirconium in ZrHx from ENDF/B-VIII.1, which contains coherent and incoherent elastic data. Inelastic cross sections differ but are generally small compared to elastic ones, except at very low and epithermal energies, where they influence the total cross section. Thermal scattering data were also generated for zirconium in ZrHx, but their inclusion had minimal effect on the benchmark results. Validation was performed using 15 benchmarks from the ICSBEP handbook for with the MCNP inputs are available. Overall, the JSI data from this study reduce the spread in the benchmark results.

The term “nuclear security” has been gradually interwoven into the Slovenian legal system since 2013 or so while “physical protection” – one of the key elements of it – has been firmly taken into consideration for three decades and beyond.

The year of 2024 was outstanding for a wealth of reasons and the international dimension of nuclear security has entailed quite some additional activities, engagements and attention – paid by the Slovenian Nuclear Safety Administration (SNSA) and other Slovenian stakeholders, like the Ministry of Foreign Affairs and the Ministry of the Interior.

In spring 2024, at the premises of the International Atomic Energy Agency (IAEA) in Vienna, the fourth International Conference on Nuclear Security (“ICONS”) brought together politicians, technical experts and others to discuss different aspects of nuclear security. The flamboyant title and recurring motto of the conference was “Shaping the Future” and it really encompassed the current overview of the activities of the Member States as well as IAEA, including their forthcoming endeavours.

Till the next ICONS in four years or so, also the second review conference for the Amended Convention on the Physical Protection of Nuclear Material (A/CPPNM) will take place. This foundational treaty is of vital importance and holding such periodic reviews could build upon trust, enable sharing of national approaches and creating the opportunities for countries to make further commitments. This goes along with the promotion of the universalisation of the treaty.

In autumn 2024, the SNSA eagerly hosted the annual – plenary meeting of the European Nuclear Security Regulators’ Association (ENSRA). It was the first meeting, held in Slovenia, of this association of the European partner (and like-minded) regulators, responsible for nuclear security. Since the detailed outcomes of the discussions are not public, for obvious reasons, the article will summarise non-sensitive pieces of information which will unveil quite some benefits of such regional associations and gatherings.

One of the most recognised and appreciated tools of the IAEA is also the Incident and Trafficking Database (ITDB). Despite being a voluntary mechanism, it continues to serve as a valuable resource in exchanging information regarding a variety of incidents of nuclear and other radioactive material out of regulatory control. Slovenia – being a member since 1995 – will continue to be active, caring und competent partner. The milestone of the 30th anniversary of the ITDB in 2025 gives an exploitable opportunity to assess the benefits of it as well as scanning further avenues of outreach and improvement.

Based upon the current (IAEA’s) plans, the next couple of years – till the second quarter of 2029 – will be marked with a revision of four top-tier documents from the Nuclear Security Series (NSS). The process will duly be followed also by the Slovenian stakeholders. Even though that no “tectonic”, text-related changes are envisaged in this revision process, a colossal kind of work will be necessary, globally, to meticulously look on all those recommendation-level documents which are the prime benchmarking “tool”, e.g. during the international peer review missions.

International initiatives (like INFCIRC/908, 910 and 918), the WINS (World Institute for Nuclear Security) as well as the new forum “Global FTPRNT” (the Global Forum to Prevent Radiological and Nuclear Terrorism) are additional platforms that bring together like-minded partners and spur up new documents or enable a fruitful sharing of good practices.

It is up to the key Slovenian stakeholders to address the path walked in the last couple of years and think about the continuous improvement in different aspects of nuclear security. The regular tasks are diverse and require human resources with competences. They span from the revisions of physical protection plans for nuclear facilities and transports of nuclear (fissile) material (based upon the regulation amended in 2023) to following and managing the process of a nuclear threat assessment, the work and mandate of the appointed Commission on the Physical Protection of Nuclear Facilities and Nuclear and Radioactive Material and so on. As in many other countries, also the security of radioactive material – including during transport – has gradually received more attention. One of the recent engagements is also the process of periodic safety review (PSR) – having in mind “physical protection” as one of the newer elements to holistically address nuclear safety of nuclear facilities, having also “security strand” or safety-security interface on board.

The article intends to capture more than just a tip of nuclear security, balancing both top-down as well as bottom-up approaches. The vivid, tireless work of the Slovenian stakeholders and keen individuals represent a small but important piece of global efforts – aiming at building a safer and more secure world and protecting against the threats in the sphere of nuclear security – reaching from nuclear facilities to transport of high-consequence radioactive material.

World energy is expected to continue increasing 2.25%/a through 2100, requiring ~40 terawatts-electric average generation by 2105, equivalent to 120 terawatts-thermal. Sufficient utility-scale energy storage to average 40 terawatts wind and solar energy, ~2 terawatt-a, costing ~2000 trillion USD at 0.10 USD/Wh, will never exist. Wind and solar energy collection cost for ~500 TWe nameplate will add another ~1500 trillion USD, not counting ~500 trillion USD transmission cost. Absent utility-scale energy storage, wind, solar and big hydro will never average more than 3 terawatts electric generation. CO2 mitigation requires atomic power expansion 5%/a starting 2025 to 50 TWe nameplate. Otherwise fossil fuel depletion achieves ~1300 ppm CO2 by 2100. Atomic power can be any combination of: (1) seawater-uranium-fuelled LWR, (2) FBR, or (3) CANDU D2O slow-neutron pile. Sufficient D2O will be available from electrolysis and fuel-cell hydrogen consumption. Atmospheric CO2 modelling assumes ocean continues absorbing 1/3 of industrial CO2 emissions. Fossil fuel is modelled as gasoline, C8H18. Maximum CO2 is ~850 ppm around 2110. After fossil fuel is phased out, CO2/CH4 (GHG) atmospheric half-life is estimated 83 a, resulting in 350 ppm CO2 around 2350. Results presented in this work are a modification of 2016 Korean prize-winning paper that used now-obsolete 2020 atomic power deployment.

Mauna Loa 2023 and 2024: +2.25%-CO2/a: 50 year epic environmental policy fails.

280 ppmCO2-1800 + Exp(0.0225/a(2025.3a - 3a[COVID19 & 2008 recession] - 1800a))

= 428.7 ppmCO2 versus April 1, 2025 Mauna Loa: 427.5 ppmCO2.

September 02, 2022, CA Governor Gavin Newsom signs CA SB No.846, Dodd. "Diablo canyon power plant: extension of operations."

28 April 2029 10:33 GMT Iberian Peninsula ~noon blackout: 15 GW (~1/16 load) from France lost. Wind and solar were providing ~25/32 Peninsula load.

Alvin M. Weinberg, Figure 3 Nagasaki data, Physics Today, March 1981:

Graph does not show significant mortality risk, absent modern medicine, below 2 Sv gammas.

December 2019 watched only rooftop PV installed on my street. Three Stooges scene: Worker took off hard hat; Battery drill rolled off roof and missed by one meter.

0-54-0 fertilizer is presently made by reacting phosphate rock with H2SO4. Most H2SO4 is from sulphur extracted from high-sulphur oil. Absent oil, making 0-54-0 will require as much power as making aluminium: ~2% world electricity. By-product uranium from 0-54-0 production could supply all world energy until 2100.

"Energy conservation" [First Law of Thermodynamics: Energy always conserved] worse than atomic power: 14 June 2017 UK Grenfell insulation retrofit fire killed more than total known deaths at Chernobyl, Fukushima, TMI and Chalk River. Jimmy Carter, Chalk River prompt-critical recovery, died at 100.

A control over radioactive material which is indivertibly present in cargo and goods is of a special concern as if unnoticed it might lead to severe risk to a general public and workers or it can cause contamination of environment. In last decades Slovenian Nuclear Safety Administration (SNSA) developed well established system related to such situations in Slovenia which is based on the SNSA on-duty officer available 24/7. The system first addressed recycling industry in country using metals originating in Slovenia or abroad. Later the systems was enlarged addressing other flows of materials where radioactive materials might pose a risk.

The SNSA database of all events with such materials where involvement of the SNSA inspectors was needed, was established in 2002. It enables analysis of different types of events as well as trends in such events. It can be used for learning purposed.

From 2021 the control of goods includes a control of packages at the post offices. Various items with radioactive materials are sent by the post and include compasses with radioactive paint and thoriated lenses as well as various other objects triggering alarms. A comprehensive analysis of risk when handling such materials by a member of the public or a worker was conducted by Jožef Stefan Institute to support inspection decisions. The SNSA decisions on a use of such items once discovered, is presented and challenges discussed. International approach including EU to managing such items present in international trade is also discussed.

The decarbonisation of energy generation is the main concern to fight climate change. Limiting the average temperature increase to less than 2°C above preindustrial levels will require economic and social sacrifices. Global energy-related and industrial process CO₂ emissions are only projected to slowly fall, so climate change will remain a global problem with disparate impacts across countries. According to the IAEA renewables and nuclear energy will dominate the growth of global electricity supply over the next years, together meeting on average more than 90% of the additional demand. The renewable energy sources and nuclear energy are often called „zero carbon”, but if the whole life cycle is considered (Life Cycle Assessment or LCA), even these low-carbon sources can contribute to significant emissions. Although they have at least one order of magnitude lower emission than the fossil options, LCA carbon intensity of the now-carbon technologies depend very much on the exact choice of the technology and the emissions related to the used electricity storage. A new nuclear option, the small modular reactor (SMR) is becoming realistic option in the second half of this decade. While many papers examined the carbon intensity of the contemporary options our paper is aiming to compare the LCA emission of SMRs and the renewable energy sources with special emphasis on the contribution related to energy storage. If Li-ion batteries or hydrogen storage is used, the LCA carbon emission of the renewable electricity can exceed 50 g/kWh while SMRs do not need energy storage and many technical options are capable for load-following operation, what strongly reduce their LCA emission.

On October 4th, 2023, a reactor coolant leak occurred at the Krško NPP which required a manual plant shutdown to determine leak location and forced outage to implement appropriate corrective measures. The Slovenian Nuclear Safety Administration (SNSA) was promptly informed of the event and initiated immediate detailed inspections and maintained continuous regulatory oversight throughout the unplanned outage.

Upon the discovery of the leak's location on October 8th within the safety injection system pipeline, the SNSA propose Krško NPP to conduct inspections of safety class 1 snubbers and a visual examination of their attachments on the SI pipelines, demonstrating proactive regulatory engagement.

The SNSA, supported by Technical Support Organizations (TSOs), provided comprehensive oversight of all repair activities, with a primary focus on ensuring nuclear and radiation safety. This included the supervision of all repair work by the SNSA inspectors and joint oversight of forced-outage activities, performed by the SNSA experts with the support of TSOs. This oversight encompassed the review of design documentation, the results of inspections, and the organization of technical meetings, highlighting a thorough regulatory process.

The power plant restart was authorized by the SNSA and TSOs solely upon the successful completion and confirmation of all prescribed tests, emphasizing the regulatory authority's role in safety verification. Furthermore, the SNSA interacted with the Belgian regulatory authority concerning the Doel event, indicating a commitment to learning and international collaboration. A comprehensive analysis of the event, including corrective actions, is being conducted by the SNSA, demonstrating their commitment to continuous improvement and safety enhancement.

Since 1966, the Jožef Stefan Institute (JSI) has been operating a 250 kW TRIGA research reactor, with continuous attention to its safe and efficient performance. For more than a decade, Safety Performance Indicators (SPIs) have been systematically tracked. Key indicators include reactor operating time, the number of irradiated samples, radiation doses received by the operational staff, and the release levels of radioactive gases into the environment. This paper presents and evaluates the SPIs recorded in 2024, offering insights crucial for maintaining and improving the reactor’s long-term safety and reliability.

In the field of research, we have continued most of the established research campaigns from previous years. Measurements were performed on the water activation loop setup (KATANA), in collaboration with the Czech Institute in Řež we tested stilbene-based detectors for neutron detection, researchers from CEA conducted experiments involving a large fission chamber and team from Lancaster University carried out characterization of a collimated radiation detector. Throughout the year, irradiation of various samples for CERN continued and we also initiated experiments in the field of BNCT (Boron Neutron Capture Therapy) in collaboration with Faculty of Chemistry and Chemical Technology.

As part of educational activities, numerous exercises were performed out by students from the following universities: University of Ljubljana, Uppsala University, Aix Marseille University, Politecnico di Milano, Chalmers University and King Fahd University of Petroleum. Besides that, one ENEEP course was organised in the end of June where participants were young professionals in the nuclear field. We also conducted a shorter version of Nuclear technology course for Slovenian Nuclear Safety Administration personnel. In March, an Open Day event was organized to educate the general public about the operation of the TRIGA reactor, attracting around 800 visitors.

In the operation of the reactor, we successfully addressed several technical challenges. A leak was detected in the secondary cooling circuit and the cooling tower. The secondary circuit was repaired, while the issue with the cooling tower was bypassed by connecting to the hydrant network, ensuring continued safe operation. Additionally, we resolved the issue with the instrumented fuel element in which all three thermocouples had failed in 2022. All thermocouples were repaired, and the element was returned to full functionality.

In the context of the European EVEREST project (Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT, GA: 101163288), the primary objective of this work package is to produce dedicated experimental data to validate advanced multiphysics models for reactor physics. The project aims to demonstrate the usefulness of advanced multiphysics simulations for long-term monitoring of pressurized water reactors, specifically focusing on reactor pressure vessel fluence calculations. Ultimately, EVEREST seeks to improve the safety margins and lifetime of nuclear reactors by reducing uncertainties, contributing to climate-neutral energy production.

To validate the multiphysics calculation models, measurement campaigns are planned at JSI’s TRIGA research reactor, with the goal of obtaining high-accuracy experimental data via non-invasive techniques. In particular, this work aims to obtain spatially precise water temperature and density data for the TRIGA’s tank and core regions.

Besides the use of thin, fast-reacting thermocouples, the possibility of using several arrays of ultrasonic transducers is being studied in the present work. The high frequency of ultrasonic elements allows for fine spatial resolution in the measurement of water temperature and density inside the tank of the TRIGA reactor. This, together with the characterisation of neutronic parameters, will provide a complete set of data for the validation of the calculation models.

At the current stage of the research project, the technique's feasibility is being assessed through several simulation methods. Some promising results were obtained through the full-wave propagation Python simulation package k-Wave, which shows that the sound speed in water can be precisely measured in a two-dimensional section of the water tank via time-of-flight calculations, allowing for an estimation of its temperature and density profiles.

To obtain significant precision in the simulations, the number of transducers has been set to the order of 32 to 64, spanning a circle of 2 meters in diameter. A study of the possible emission frequencies has been performed, and 40 kHz transducers were preliminarily chosen, as wave attenuation at the required distances becomes too significant with higher frequencies. Furthermore, 40 kHz elements are widely available commercially, facilitating potential implementation in other research reactors. In addition, the possibility of obtaining precise 3D measurements using multiple circular arrays will be studied.

A simplified approach has been started in parallel, using first-arrival travel-time simulation and reconstruction through the Python package PyGIMLi. Although it does not compute full wave propagation or account for the effects of echoes, the produced reconstructions also allow for distinguishing regions of different water temperatures satisfactorily, while being less computationally costly. A reconstruction with the inclusion of a circular obstacle (representing steel internals on the TRIGA tank) was also achieved.

In a separate line of investigation focusing on the core region, also using k-Wave, a dedicated three-dimensional simulation has been initiated to study ultrasound wave propagation along a single fuel element placed inside the reactor core. The simulated fuel element measures 80 cm in length and 4 × 4 cm in cross-section and incorporates realistic material density variations to represent the internal heterogeneity of the structure. Two transducers emitting at a central frequency of 40 kHz are positioned along the element to study wave propagation in the axial direction. This high-frequency setup allows for the simulation of detailed pressure fields and the tracking of complex reflection patterns within the fuel element geometry. As the simulated pressure field is directly related to local variations in temperature and density, this approach enables a realistic assessment of the sensitivity to changes in physical conditions within the reactor core. By modelling both the direct transmission and internal reflections of the acoustic waves, the aim is to evaluate the feasibility of extracting meaningful thermal information from the fuel elements.

Once the most optimal configuration has been determined through the preliminary simulations, an experimental trial is envisioned for the second half of 2025. It will start with a benchmark model designed to evaluate the system’s ability to detect subtle temperature variations in water.

In conclusion, this work will set off part of the experimental campaign of the EVEREST project at JSI, providing experimental data for the validation of advanced multiphysics models, and potentially creating a novel methodology for precise in vivo measurements in research and industrial reactors.

High-Performance Computing (HPC) involves the use of powerful computing systems to solve complex scientific and engineering problems that exceed the capabilities of standard computers. At "Jozef Stefan" Institute (JSI) at Reactor Center Podgorica (RCP), four on-site compute clusters have been established over the past almost two decades and have positioned it as one of the HPC centers in Slovenia. These high-performance compute clusters (HPCCs) play a vital role in nuclear science, supporting research in neutronics, fusion, thermal hydraulics, materials science and structural integrity.

Reactor Physics (F8) and Reactor Engineering (R4) and departments have jointly operated and continually, annually upgrade their HPC infrastructure with support from the Slovenian Research and Innovation Agency (ARIS). These clusters are essential not only for advancing research quality and nuclear safety but also serve as a valuable tool in education and training of students and young researchers.

An HPCC combines advanced hardware—including multi-core servers, high-speed networks, and reliable storage—with software systems, running both commercial and open-source applications. Increasing computing demands have led to infrastructure improvements at RCP, including upgrade of the cooling system and new server room monitoring system. The most significant improvement in energy efficiency is the use waste heat, captured from the HPC cluster to heat the TRIGA reactor hall. All available data from the server room is aggregated and driven with Home Assistant process information system.

This study presents high-precision thermal mapping of the TRIGA Mark II reactor pool, conducted within the framework of the EURATOM-funded Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT - EVEREST project (GA №101163288). Building on previous experimental efforts, the current work significantly improves both the spatial and temporal resolution of temperature measurements within the reactor core and the overlying pool region. These improvements are achieved through the deployment of a greater number of smaller thermocouples.

Beyond temperature profiling, the enhanced sensor network enables the analysis of turbulent structures and flow characteristics by examining correlations between sensor signals. This provides indirect but valuable insights into local coolant velocity and turbulence intensity in the reactor pool environment.

Attempts to characterize the flow field using optical methods are also under development, utilizing the Background Oriented Schlieren (BOS) imaging technique, where the flow is discerned from the image flickering due to the changes in water’s refraction index due to temperature changes. Initial tests in the JSI TRIGA reactor were performed in the core using a large background pattern, oposite of which a GoPro camera was located in close proximity to the core. The test was not sucesfull due to the:

l Camera location in close proximity to the reactor core, which shut down the camera at a power lever of ~10 kW.

l Multiple obstacles between the camera and background such as s control rod mechanisms and irradiation channels.

Due to the above-mentioned issues, we’re currently investigating the possibility of projecting the background pattern using a laser diffraction grating, amplyfiying the effect and allowing for projection on the intermediary structures, while allowing for a larger distance of the camera from the reactor core. The aim is to place an array laser-camera pairs on the reactor edge’s perimeter, which would allow for a tomographic resonstruction of the flow field.

The resulting high-resolution, benchmark quality datasets will serve for validation of advanced multi-physics simulation tools—one of the central objectives of the EVEREST project and will be made available for a wider audience via the OECD NEA repositories trough Expert Group on Reactor Systems Multi-Physics (EGMUP) and Task Force on Artificial Intelligence and Machine Learning for Scientific Computing in Nuclear Engineering initiatives.

Safety and Relief Valves (SRVs) are used in light water reactors to limit the pressure in the reactor coolant system and to relieve pressure in case of incidents and accidents. While these valves are considered to be very reliable under design conditions, their behavior beyond these conditions is largely unknown. In a postulated high pressure core melt accident, an early failure in open-position, fostered e.g. by the repeated opening cycles, high gas temperatures or entrained particles, could largely influence the accident outcome, as it prevents less favorable depressurization scenarios and determines the path of radionuclides and hydrogen. Investigations in the aftermath of the Fukushima Dai-ichi accident have highlighted this aspect and led to its inclusion in recent accident related Phenomena Identification and Ranking Tables (PIRTs). However, safety and relief valves are complex components with a variety of designs, rendering a generic and accurate prediction of the failure behavior difficult.

In the present work, a staged approach is adopted to better understand and predict the failure behavior of the most used pilot-operated and spring-loaded valve types. Besides a brief description of the basic functional principles, potential failure modes are summarized, evaluated and systematized into the clusters ‘stochastic failure’, ‘fluid damage’, ‘temperature induced weakening of the structure’, ‘jamming’, ‘pilot and actuation failure’ and ‘setpoint drift’. This is based on a literature review of previous work, a comprehensive evaluation of operating experience (under non-accident conditions), related problems in steam power plants and recent exchanges with valve experts and manufacturers. A key finding is that pilot-operated relief valves are suspected to be primarily susceptible to jamming or pilot-valve failure. In light of the failure modes discussed, a failure temperature of the structure in the range 600-800 °C seems to be a plausible assumption. Spring-loaded valves are expected to be much more robust but may be prone to setpoint drift due to temperature-induced weakening of the spring.

Based on this knowledge, numerical studies are carried out. This includes a Computational Fluid Dynamics (CFD) simulation of the flow through the valve using results from the AC2 accident code package as boundary conditions at the fluid inlet and outlet. For the case studied, the simulation gives no indication of entrapment or severe erosion potential of entrained particles. Furthermore, the temperature gradients at the most vulnerable points are relatively low. Most of the pressure loss and fluid expansion is concentrated around two tapers with critical flow in the outlet of the valve. The complex flow pattern with very high velocities (Mach 2.7) ensures a high heat transfer from the fluid to the structure, so that the heating of the structure is mainly determined by the thermal inertia of the parts. Another numerical study involves a Finite Elements (FE) calculation of the temperature field of the structure, which is the main factor influencing various failure modes, such as jamming. It shows temperature differences due to different thermal inertias of differently shaped parts. From the temperature distribution, the evolution of the gap width between piston and cylinder, e.g. by thermal expansion, is predicted.

Finally, potential experimental work is discussed. Considering the huge cost of implementing full-scale mock-ups for each valve type and the stochastic nature of some effects requiring multiple test runs, single effect tests or tests on parts of the valve appear to be a cost-effective first step. With this in mind, a pilot study of the sticking and frictional behavior of two surfaces, resembling the piston and cylinder of a valve made of Stellite 6 and 17-4 PH austenitic steel is described. The experiments, carried out in a furnace at room temperature, 600 °C, 800 °C and 1000 °C show a significant sticking between the surfaces, albeit the force required to overcome the sticking does not increase with temperature. The friction increases from 0.2 to 0.5-0.7 at elevated temperatures due to the formation of a corrosion layer and returns to values around 0.4 after cooling.

In summary, the present work uses expert feedback, numerical simulation and experimental work to narrow down failure the modes of safety and relief valves in light water reactors during high pressure core melt accidents, to allow an informed estimation of the failure temperature and to propose a path towards a reliable failure assessment model.

Lead-cooled fast reactors (LFRs), identified as a key Generation IV nuclear technology, offer compelling advantages for sustainable energy production due to their inherent safety features—such as operation at atmospheric pressure, high thermal conductivity, passive heat removal, chemical stability, and improved fuel utilization. However, the extreme operating environment of LFRs, characterized by broad temperature gradients and high irradiation exposure, poses significant challenges to the mechanical stability of fuel assemblies and other core internals. These conditions can result in elastic and plastic deformation, affecting core reactivity and safety margins over the reactor's lifecycle.

This study investigates the thermomechanical response of single assemblies and ring-averaged regions in representative LFR cores, with emphasis on the influence of core restraint system configurations. A simplified core model is used to analyze temperature fields and across-duct temperature gradients, which are critical in validating point kinetic approximations in coupled neutronic-thermal kinetic (NK-TK) simulations under accidental scenarios. The simulation chain combines NUBOW-2D INEL for inelastic mechanical response and ATHLET for thermal-hydraulics, initially coupled via a point kinetics approximation, with provisions for integrating higher-fidelity neutronic calculations in future developments.

A key contribution of this work is the introduction of pyNubow, a flexible Python-based interface designed to streamline simulation setup, execution, and postprocessing. pyNubow ensures consistent geometric input handling, simplifies output data collection and visualization, and manages batch execution to facilitate parametric studies. The interface is designed to support integration with subchannel thermal-hydraulic codes such as DASSH, opening paths toward higher-fidelity multi-physics coupling in future studies.

While experimental validation data for LFR systems remains limited, simulation results will be benchmarked against selected IAEA 1989 reference cases. Additionally, pyNubow offers the flexibility to support future code-to-code validation efforts through coupling with finite element tools, though such comparisons may fall beyond the immediate scope of this study.

The most promising perspective advanced tolerant fuel (ATF) cladding candidates for possible application in commercial nuclear power plants (NPPs) throughout the world include:

• FeCrAl alloy cladding;
• zirconium-based cladding with protective chromium coating (Zr/Cr cladding);
• chrome-nickel alloy cladding with high Cr content;

• SiC-based ceramics composite cladding.

These materials have excellent characteristics of corrosion and oxidation resistance compared to zirconium both for the nuclear power plant (NPP) normal operation temperatures and high-temperature conditions.

In this paper, the advanced models of high-temperature oxidation of FeCrAl, Zr/Cr, Cr-Ni and SiC/SiC cladding were developed. The phenomena responsible for worsening of growing oxides protective properties (which are different for the claddings considered) were taken into account. For example, the effect of Cr-Zr interdiffusion with subsequent influence on degradation of protective properties of Zr/Cr cladding was considered.

The comparative analytical and numerical calculations of these ATF-claddings behaviour under NPP design-basis loss-of-coolant (LOCA) accidents are performed. The temperature scenario for LOCA was considered similar to QUENCH-LOCA test series conducted about ten years ago in KIT, Karlsruhe, Germany. It was found that all four ATF-claddings show excellent results: a slow oxidation kinetics and a small hydrogen production compared to standard zirconium-based alloys claddings. However, in the case of transition of design-basis-accident to beyond design-basis-accident with corresponding rise of temperature, we observe the considerable worsening of output parameters. The reasons for it are discussed in this paper.

Despite the existence of obvious mechanisms leading to loosing of protective properties at high temperatures, one can make a conclusion that the application of FeCrAl, the chromium-coated Zr-based cladding, the chrome-nickel claddings and SiC-based composite cladding is optimistic for considerable upgrade of safety level for NPPs for design-basis-accident conditions with maximum temperatures T<1200°C.

TRANSURANUS is a fuel performance code developed and maintained by the Joint Research Council (JRC) in Karslruhe, Germany. The code is used worldwide by research institutions, nuclear safety authorities and companies for the prediction of the fuel rod performance during normal and abnormal operation as well as under accident conditions in various reactor types. Therefore, a thorough in-depth validation of the code by comparing predictions with test results is of vital importance. In this context, the legacy LOCA test at the research reactor in Halden (Norway) are especially interesting.

The present study focusses on the LOCA test IFA 650-7, which was conducted in 2008 with a 48 cm long rodlet with a medium burnup of 44 MWd/kgU from the Leibstadt boiling water reactor (Switzerland). This test poses some specific modeling challenges due to the inhomogeneous pressure distribution in the test flask during the blow-down phase, which led to an uneven heat-up of the cladding, and the comparatively large free volume. Since it is not possible to directly model the blow-down in TRANSURANUS, we use the built-in option to directly prescribe the cladding temperature during the blow-down and the heat-up phase. For this purpose, we derived several cladding temperature profiles based on the measured data and thermodynamic hypotheses with varying degrees of detail. We also investigated how accurately the history of the mother rod must be modeled and whether the inner pin pressure or the temperature of the free volume need to be prescribed as well to reproduce the experimental results. While all our models predict a cladding failure, the partially large discrepancy between the predicted and the measured time of the cladding rupture does emphasize the importance of accurately modeling the cladding temperatures during the LOCA test. The rod history determines the oxide layer thickness, the hydrogen content, the cladding creep and stress, as well as the geometry, which define the initial conditions of the rodlet for the LOCA test. The inner pin pressure resulting from the history run is discarded due to the refabrication prior to the LOCA test. The neutron flux, which we estimated using the two-dimensional deterministic neutron code module TRITON from the SCALE 6.2 code system and standard BWR irradiation conditions, mainly affects the gap width and the cladding creep and stress. While moderate modeling choices in the history do certainly influence these initial conditions, they have little influence on the predicted time of the cladding failure since the increase of the cladding temperature and the inner rod pressure are the main driving forces during the LOCA test. The full paper will discuss these modelling choices in detail and compare their influence on the predictions to provide insights for the modelling of other LOCA test at the Halden reactor.

Research on containment atmosphere phenomena during severe accidents in light water reactors, mostly related to the risk of hydrogen explosion, has been going on during the past decades. As for other severe accident phenomena, the necessity of further research in this field is currently being evaluated within the European SEAKNOT project (2022-2026). Since the beginning of the project, among other results, a separate Phenomena Identification and Ranking Table (PIRT) for containment phenomena was established.

The following was accomplished first:

- a comprehensive list of possible phenomena was established;

- the experimental and the theoretical knowledge about each listed phenomenon were evaluated;

- the safety relevance of each listed phenomenon was assessed.

For all the considered aspects (experimental knowledge, theoretical knowledge, safety relevance), the following rankings could be assigned: High, Medium or Low. Assessments were based on individual expert judgement/assessment and later discussions of singular and significantly controversial evaluations. For the specific case of containment phenomena, separate evaluations and assessments were made for the "in-vessel" and "ex-vessel" phases of a severe accident. Namely, first, some phenomena are expected solely during one of the phases. Second, although phenomena supposed to occur during both phases are basically the same, not only can some boundary conditions be different (which is reflected in both experimental and theoretical knowledge), but their influence on possible nefarious consequences of the accident (which is reflected in the safety relevance) may also be different.

According to the preliminary outcomes, some interesting findings will be reported in the paper. Among them, one may highlight: no high priority issue was found, this meaning that containment has been vastly investigated in the past, although some gaps in boundary conditions of current and upcoming reactors (water-cooled Small Modular Reactors) might need further attention; some phenomena are considered highly safety relevant with still some knowledge necessary; even if a lot of knowledge has been accumulated, the level of lumped-parameter graded data and Computational Fluid Dynamics graded data are far from being even. Further discussions might yield some changes in these preliminary insights.

These abstract outlines a comprehensive study on the External Reactor Vessel Cooling (ERVC) process for maintaining the integrity of the reactor pressure vessel (RPV) during severe accidents in small and medium pressurized water reactors (PWRs), such as the one at NPP Krško. The research focuses on assessing the applicability of ERVC for PWRs that were not originally designed for in-vessel retention (IVR) by applying ERVC. Key points include:

• Theoretical and numerical assessment of ERVC applicability, including potential negative effects, design limitations, and improvement potentials.
• Reactor cavity geometry, available time for dominant plant damage states (PDS), RPV size and thermal power, and the influence of RPV reflective thermal insulation.
• Using plant-specific MAAP code calculations, operational procedures for severe accident management, technical manuals, and relevant scientific studies.
• Defining simple, effective actions for successful IVR, assessing containment flooding strategies under degraded conditions, and proposing feasible design modifications for improvement.

This study aims to provide actionable insights for industry application to enhance safety measures in PWRs during severe accidents.

The calculation of NEK Station BlackOut (SBO) High Pressure (HP) and Low Pressure (LP) scenarios without any mitigation for the first 24 hours was performed using the MELCOR code. HP and LP sequences differ only in hot leg creep break assumed in LP sequence. The RCP seal leakage was calculated without and with implementation of passive shutdown seals. The containment model includes PARs and Passive Containment Filtered Vent (PCFV) system up to the exhaust duct. Plant specific fuel source term is calculated using ORIGEN code for real plant’s operation history data. The released radiological effluents were calculated using the MELCOR thermal hydraulic results and RADTRAD containment model based on RG 1.183 assumptions. The dispersion X/Q factors needed to perform calculation of doses in the environment (Exclusion Area Boundary EAB and Low Population Zone LPZ locations, and 1D spatial dependence up to 200 km) were based on the Lagrange particle methodology developed by MEIS d.o.o. In order to verify used X/Q factors for EAB and LPZ locations, TEDE (Total Effective Dose Equivalent) doses were calculated using the RG 1.249 ARCON based atmospheric dispersion factors too. Finally, 2D distribution of TEDE dose was obtained using JRODOS calculation for selected past meteorological data, assuming duration of release of 72 hours. Due to earlier passive actuation of PCFV, higher doses are obtained for LP sequence, but thanks to effective reduction of effluents in PCFV filters, the obtained TEDE doses are still reasonable.

Advanced Technology Fuels (ATFs) were developed in response to the Fukushima nuclear accident of 2011. Although initially conceived to enhance nuclear power plant safety under accident conditions, the behavior of these materials under normal irradiation has been postulated as crucial for their potential deployment in commercial reactors. In this context, the METATF collaborative research agreement between NFQ and the Spanish Nuclear Safety Council (CSN) was launched in November 2023. The primary objective of the METATF project is to develop an advanced Fuel Safety Evaluation Methodology (FSEM), based on the combination of the TRACE and TRANSURANUS (TU) computational codes. This methodology is designed to evaluate the full lifecycle of fuel rods within the core while considering the specific performance of ATF materials.

The FSEM methodology comprises three key stages. First, the base irradiation phase is simulated using the TRANSURANUS code for all fuel rods included in the analysis. Subsequently, transient Thermal-Hydraulic Boundary Conditions (THBCs) are generated using a comprehensive TRACE model of the nuclear plant, initialized with the output from the preceding TU simulations. Finally, these THBCs are incorporated into the TU inputs to conduct a Best Estimate Plus Uncertainty (BEPU) analysis, employing Monte Carlo sampling for the whole set of fuel rods. This enables a detailed evaluation of rod performance and facilitates probabilistic estimation of cladding failure (burst) and the total number of failed rods under postulated transient conditions.

This study applies the aforementioned FSEM methodology to a Large Break Loss of Coolant Accident (LBLOCA) scenario in a generic three-loop Pressurized Water Reactor (PWR), with the objective of quantifying the total number of failed fuel rods. The configurations analyzed include UO₂/Zircaloy (Zry), UO₂/FeCrAl, UO₂/Cr-coated Zry, Cr-doped UO₂/Zry, and U₃Si₂/FeCrAl, allowing for a preliminary assessment of the benefits associated with various ATF concepts.

To perform the analysis, a full-plant TRACE model of the generic three-loop PWR was developed, featuring a core representation composed of 66 fuel rods—33 representing hot channels and 33 representing average channels. Notably, the TRACE code was modified to incorporate the physical properties and behavioral models relevant to FeCrAl and Cr-coated Zry materials. Additionally, the standard intragranular diffusion coefficient of UO₂ in the TU code was adjusted to model the performance of Cr-doped UO₂.

For the reference case employing conventional fuel materials, the Monte Carlo simulations revealed a non-negligible probability of cladding failure, affecting not only rods in the hottest channels but also some in average conditions. In contrast, simulations involving ATF configurations demonstrated a marked reduction in both cladding oxidation and fission gas release. These improvements were primarily attributed to the use of FeCrAl and Cr-coated Zry claddings, as well as Cr-doped UO₂, respectively, fuel during the base irradiation period. While only minor variations were observed in the THBCs computed with TRACE (dominated by the influence of the LOCA transient conditions and the cladding material properties), a significant decrease in the number of failed rods was evident in cases employing ATF cladding technologies.

At present, there is increasing interest in Accident Tolerant Fuels (ATFs) or Advanced Technology Fuels for light water reactors. These fuels have the potential to improve the nuclear safety and operational flexibility of commercial nuclear reactors. The ATFs incorporate advanced cladding materials and fuel compositions specifically designed to withstand higher temperatures, improve corrosion resistance and enhance fuel performance under both normal operating conditions and accident scenarios. Among the various ATF designs, FeCrAl cladding is one of the most promising. Made from an iron-chromium-aluminium alloy, it offers exceptional resistance to oxidation and corrosion at high temperatures. This makes it particularly useful in the event of accidents, where conventional zirconium-based cladding can oxidize rapidly and produce hydrogen. These reactions generate large amounts of heat and hydrogen, which can create a risk of explosion when mixed with oxygen.
To assess the safety aspect of ATFs, several analyses of typical accidental sequences in a LWR as the LOCA or the Station Blackout are performed using TRACE and MELCOR system codes. In particular the FeCrAl and Cr-coated claddings are used as the ATF fuel which will employ the oxidation correlations obtained during recent experiments. An assessment for a generic PWR Westinghouse. The main differences of using ATF will be exposed relative to regular fuel in terms of timings, temperatures and damage conditions along the accidental sequence. Besides, core damage frequency is quantified by means of a RiskSpectrum model for a generic PWR.
The simulations show that the performance of the ATF in different accidental scenarios and the quantification with RiskSpectrum allows to obtain the risk reduction. This study shows that the increase of available times during accidental sequences and the risk reduction (5%-25%) due to different ATF fuels.

The development of methods and tools for uncertainty quantification and sensitivity analysis is still ongoing. The objective of this study was to test the capabilities of the latest SNAP (Symbolic Nuclear Analysis Package) software and its uncertainty plugin, which uses DAKOTA (Design Analysis Kit for Optimization and Terascale Applications). For the uncertainty and sensitivity analysis, the LOFT L2-5 test was used, which was double ended guillotine large break loss-of-coolant accident (LOCA), conducted on the Loss of Fluid Test (LOFT) experimental facility. The LOFT facility was a pressurized water reactor (PWR) with a thermal power output of 50 MW and two loops. It was designed to study the thermal-hydraulic response of the system to various simulated LOCA scenarios and therefore included systems not found in commercial reactors, specifically for testing purposes. For calculations the latest TRACE V5.0 Patch 9 thermal-hydraulic computer code was used. TRACE input model of LOFT L2-6 test in ASCII form, developed by the U.S. Nuclear Regulatory Commission for evaluating TRACE version V4.160 from 2005, was used. LOFT L2-6 test represented a loss-of-coolant accident (LOCA) under different initial conditions than those of the LOFT L2-5 test. Therefore, in this study the input deck was modified to LOFT L2-5 initial and boundary conditions, and new steady state was calculated.

For the uncertainty and sensitivity analysis, the latest SNAP uncertainty plug-in was used, available at the time the study was performed (version 4.1.3, September 2024). The first order statistics using Wilks formula obtained by Wald, which can be used for quantification of more output uncertain parameters in the same analysis, was used.

In the study, 434 simulations were performed varying 16 uncertain input parameters, which allowed to quantify one-sided uncertainty of up to 15 output uncertain parameters.

Nevertheless, in the study 9 output uncertain parameters were considered. The results of uncertainty quantification showed that the scalar value of the highest cladding temperature (considering all fuel rods temperatures at all time steps) was enveloping the experimental value of peak cladding temperature (PCT). The results further indicate that there is a 95% probability that the maximum calculated cladding temperature will remain below the experimental value of PCT. The sensitivity analysis using Pearson and Spearman correlation coefficient showed that the most influential parameters were specific fuel capacity and fuel thermal conductivity. Finally, it was demonstrated, that SNAP Uncertainty Plug-in is efficient tool for uncertainty analysis as one job and Microsoft Word file is generated, containing output results.

A performance-based approach is one that establishes performance criteria and calculable results as the primary basis for decision-making. In case of fire protection engineering the results of the experiments or detailed fire protection calculations can provide us with fire conditions (temperature distribution, heat flux and smoke concentration) which can be used to evaluate potential for damage/unavailability of the equipment, fire propagation and spreading. In this paper FDS CFD code was used to perform calculation for two similar Krsko NPP safety related pump rooms having simple geometry and rather low thermal loading (almost the same equivalent fire severity according to Fire Hazzard Analysis). The idea is to demonstrate fire modeling procedure, required input data and potential benefits of performing such calculations. The one of selected rooms is the room with centrifugal charging (CS) pump belonging to CVCS system. The other one is room with safety injection (SI) pump. Both rooms, in first approximation, have rather limited communications with other areas what makes calculation easier (fire rated door and ducts without fire damper). Difference between those two rooms is that SI pump room has safety related safety related HVAC system inside the room, and HVAC system for CS pump room is facilitated from outside the room. Combustible inventory in both rooms is lube oil, electrical equipment and cable insulation. Fire scenario is modeled with the pump lubricating oil as ignition source. The pump skid is surrounded by a dike designed to contain lubricating oil that may leak or spill. The fire occurs following an accidental release of pump oil and ignition of oil that leaked into the dike area. In one scenario, the heat input is modeled using prescribed time dependent heat release rate (HRR), and in other scenario using lube oil chemical composition to estimate burning rates. The room door is normally closed (lower opening is modeled), but it is modeled opened in one scenario 10 min after ignition by the fire brigade. In order to assess the behavior of the forced ventilation systems on fire propagation, also different scenarios are modeled, in SI pump room one scenario assumes HVAC operation and another with inoperable HVAC. The goal of the calculation is to model heat transfer by convection and radiation between the fire and the exposed surfaces, that is fire development and related spatial influence to be able to assess the possible fire damage of electrical cables and equipment present in the room (target of the calculation, temperature damage criteria) and the temperature at fire detector position. There is no automatic fire suppression system in the rooms. The smoke distribution and soot density is calculated too. Fire modeling with calculable results provides more flexibility in achieving established performance criteria during all phases of nuclear power plant operations.

The risk of combustion in a pressurized water reactor (PWR) during a severe accident has been confirmed directly by the accident at the Three Mile Island nuclear power plant (NPP) in 1979 and indirectly by the accident at the Fukushima Daiichi NPP in 2011. These energetic events might jeopardize containment integrity, so that fission product would find their way to the environment. How to manage a severe accident to reduce hydrogen combustion risk is thus an issue in nuclear safety.

Within the European AMHYCO project (2020-2025), the risk of combustion of H₂ and CO in containments of European PWRs (Western-type, KWU-type and VVER-type) was investigated. First, simulations with system severe accident codes were performed to obtain mass and energy release rates from the reactor coolant system towards the containment. In parallel, experiments were performed to determine the recombination rate of Passive Autocatalytic Recombiners (PARs) as well as the flammability limits of H₂-CO mixtures at various pressure and temperature conditions. Using generic containment models, based on European designs, the containment responses were then simulated with lumped-parameter codes, three-dimensional codes with subgrid-scale modelling and computational fluid dynamics codes. The simulation results indicated whether in some regions of the containment, the flammability limits were exceeded.

To investigate the influence of accident management, simulations were performed by varying both the use as well as the performance of containment safety systems: PARs, sprays and fan coolers. Some of the main conclusions, specifically related to the combustion risk, are the following:

• Back-fitting of PARs has a significant benefit concerning the mitigation of possible in-containment combustion.
• A filtered venting system is not designed as a combustible gas control system and should not be treated as such in the plant Severe Accident Management Guidelines (SAMG).
• Spray operation after recovery power in a station blackout should be delayed for some time. This grace time would allow PARs to reduce significantly combustible gas concentration, so that when steam is removed from containment atmosphere, the gas mixture still stays outside the flammability domain.
• To monitor the containment gas mixture flammability, the quantification of oxygen content can be equally important as the detection of combustible gases.
• Simulations on different length scales gave an insight into how point-based measurements of the containment atmosphere can be interpreted by the main control room.

Large break loss of coolant accidents (LBLOCAs) with fast initial cooling rate and low downcomer temperatures combine to produce a high-severity transient for pressurized thermal shock. The main objective of this study is to perform advanced safety analysis of hot leg LBLOCA with uncertainty and sensitivity analysis using latest U.S. Nuclear Regulatory Commission computer codes. For analysis two-loop pressurized water reactor will be used. The calculations will be performed with the TRAC/RELAP Advanced Computational Engine (TRACE) computer code. Four uncertainty analysis Symbolic Nuclear Analysis Package (SNAP) Uncertainty Analysis plug-in using the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) will be used.

The SNAP Uncertainty Analysis plug-in employs a method for computing sample sizes based on the Wilks method. The multivariable approach of the Wilks formula obtained by Wald will be used, which allows the number of dependent figures of greater than one. The main sources of uncertainty in TRACE thermal-hydraulic code will be considered based on the literature review: the code or model uncertainties, and the uncertainties of nuclear power plant data. To each selected input uncertain parameter, the probability distribution will be assigned. SNAP Uncertainty plug-in will be used to generate random variates for TRACE input model. The variates will then be applied to the identified input parameters and the input models will be generated and TRACE computer runs performed. The obtained runs will be used for both uncertainty and sensitivity analysis. In the uncertainty analysis three figures of merit will be used: (a) primary pressure, (b) liquid temperature below reactor pressure vessel (RPV) cold leg (CL) inlet, and (c) wall temperature below RPV CL inlet.

The uncertainty results will include confidence levels, response function probability distributions, and other properties specified in the initial setup. For sensitivity results the correlations computed by DAKOTA will be included. Correlations are broken down by response function and listed as a table of simple, partial, simple rank, and partial rank correlations relative to variates and other response functions.

This article presents a discussion on comparisons of results from an analytical study of the transient process "Steam Generator Tube Rupture" (SGTR) for VVER-1000/V320 units at the Kozloduy Nuclear Power Plant (NPP).

The investigated scenario is guillotining double ended steam generator (SG) tube rupture at the upper part of the tube bundles on the side of the cold collector in the SG #1. The main features of such accident in comparison to others is that the operators have to take actions after analyzing symptoms and to manage the accident correctly. It is mean no automatic activation of reactor SCRAM.

The main objective of this study is to evaluate the possibility and possible effects in reducing time for reactor SCRAM activation as well as another operator actions based on the operator strategies from symptom-based emergency operating procedures (SB EOPs). For this purpose, the amounts of primary coolant water that will pass from the primary to the secondary side will be compared using different time schedules in the accident progression. The other objective of this study is to assess possible problems after the operator's actions that could lead to the release of radioactivity into the environment. Also, to estimate the ability for reducing primary to secondary leakage and indication for an effective strategy for preventing radioactive leakage to the environment.

For the purposes of the study, two calculations were performed with the computer code RELAP5/MOD3.3. The “base case” calculation is based on published SB EOPs analytical analyses, when the operator applies the corresponding relevant procedures. In the second scenario, the possibility of earlier application of the relevant procedure is estimated. The main aim for choosing the second scenario is to analyze the possibility for reducing the radioactive coolant leakage from primary to the secondary side.

In the investigated scenario are applied the following steps based on the SB EOPs:

• Activation of reactor SCRAM, isolation of a turbine generator and opening of the steam dump to the condenser regulators (BRU-Ks) for supporting secondary side pressure in the main steam header (MSH);
• Isolation of damaged SG#1 from feed water (FW) and switching off of main coolant pump (MCP) on the damaged loop;
• Isolation of damaged SG#1 by fast acting isolation valve (FAIV).
• Continue to cooldown of primary coolant with maximum allowed velocity of 60 ^°C with BRU-Ks;
• Primary side depressurization by using pressurizer spray;

The calculations are performed until the primary and secondary side pressures are stabilized on requested levels.

Accurate prediction of steam condensation phenomena is essential for the development and licensing of Small Modular Reactors (SMRs), which often rely on passive safety systems such as safety condensers (SACO). The need for model development and validation of passive systems with SACO has also been recognised in several SMR initiatives and projects, including the EU Industrial Alliance on SMRs and the Nuward concept.

This study investigates the implementation and performance of condensation models in system-level thermal-hydraulic codes, with a focus on the German code AC²/ATHLET. The objective is to assess the applicability of the condensation model for simulating SACO operation under representative conditions.

The investigation is based on experimental data from the in-Pool Energy Removal System for Emergency Operation test facility (PERSEO) and the Primary Coolant Loop test facility (PKL), both of which provide relevant benchmarks for SACO behaviour under various thermal-hydraulic transients. The analysis focuses on the formulation of the condensation model in AC²/ATHLET, with emphasis on the representation of heat and mass transfer during phase change in the governing equations. Furthermore, a modelling approach implemented in AC²/ATHLET is compared to other widely used system codes, like TRACE and RELAP, highlighting key differences in condensation modelling approaches and their implications on predictive accuracy. Emphasis is placed on identifying limitations and strengths of the AC²/ATHLET implementation in capturing the physical behaviour of SACO units.

Boiling is an efficient mechanism for heat removal and is being investigated for use in the divertor cooling channels of future fusion reactors, where extremely high heat fluxes must be managed. Accurate visualization of flow boiling under such conditions is essential for understanding the underlying physical phenomena and for the development and validation of advanced numerical models. To support this, the Fusion Experiment for Divertor Optimization Research Applications (FEDORA) was designed. The experiment is equipped with state-of-the-art instrumentation to observe and quantify boiling processes across a broad range of heat fluxes and flow conditions.

The complexity of such experiments requires a robust and synchronized data acquisition and control system capable of handling large volumes of high-frequency data from various sensors and transmitters. To meet these requirements, a Python-based interface was developed using National Instruments (NI) data acquisition hardware. Python was selected over commercial platforms such as LabVIEW due to its open-source nature, ease of use, and broad adoption in the scientific community. It enables direct integration of data acquisition with signal conditioning, real-time monitoring, experiment control, and even embedded image processing and visualization tools.

The developed system ensures safe operation, noise reduction, and flexibility for future extensions. A graphical user interface (GUI), built with Python’s GUI libraries, allows real-time display of signals and system status, while the underlying code handles synchronized data logging and device communication. The interface serves not only as a control platform but also as a foundation for pre- and post-processing routines, enhancing the overall efficiency and transparency of the experiment workflow.

This work demonstrates the feasibility and advantages of a Python-based approach to managing complex thermal-hydraulic experiments—not only in fusion research, but also in experimental fluid mechanics in general—providing a modular, cost-effective, and highly customizable alternative to proprietary software solutions.

Flow boiling is an important phenomenon, in particular in nuclear fission power generation. It is a highly effective way of removing heat from a surface, and its understanding and prediction is important for nuclear safety and the efficiency of energy production. This is why we are committed to developing numerical models that can accurately describe boiling phenomena. To validate such models, we need high-quality experimental data. Present work focuses on a computational fluid dynamics (CFD) model of the THELMA (Thermal Hydraulics Experimental Laboratory for Multiphase Applications) flow boiling experiment at Jožef Stefan Institute. The experiment is set up as a double pipe heat exchanger with an inner copper tube with an outer diameter of 12 mm, mimicking the dimensions of a fuel rod in the reactor core, encased in a glass tube to form a 2 mm wide annular gap for the coolant flow. The copper tube is heated from the inside by hot water flow and its outer surface is cooled by a refrigerant flow (R245fa) in the gap. Due to the transparent test section, the boiling of the refrigerant on the copper tube can be recorded and images processed to determine the bubble size distribution and void fraction. These data are complemented by thermocouple temperature measurements on the inner wall of the copper tube and in the hot water flow inside the copper tube.

The present simulations consider only the fluid domain with the refrigerant flow in the annular gap, treating the glass and copper walls as boundaries, disregarding the inner hot water flow. The CFD model represents a vertical experimental setup and, for simplicity, considers a very small three-dimensional slice of the annular gap assuming axial symmetry. Simulations were performed with the multiphase Euler-Euler model in the open-source software OpenFOAM. Firstly, simulations of single-phase heat transfer cases were conducted to investigate the effects of different turbulence models on the heat transfer in the refrigerant. Secondly, a two-fluid simulation of boiling of the refrigerant in the gap was carried out to compare the effects of different boiling model parameters.

Two complementary validations were performed. Our first objective is to use the heat flux derived from experimentally measured temperature in the flow boiling experiment as a boundary condition to validate the measured wall temperatures with the simulation. On the contrary, our second objective is to use the measured wall temperatures as an imposed boundary condition in the simulation to compare the simulated heat flux with the experimental one.

Accurate prediction of critical heat flux (CHF) is essential for the safe and efficient design of thermal-hydraulic systems. While extensive research has been conducted for pressures up to a reduced pressure of 0.7, which marks the operating pressure for state-of-the-art pressurized water reactors, and for pressures above the critical point in the context of an growing interest in supercritical reactors, a notable gap remains in the transitional pressure regime between a reduced pressure of 0.7 and 1. Although this range may not represent steady-state operating conditions, it becomes highly relevant during start-up, shut-down, and accident scenarios where supercritical systems can transient through very high subcritical pressures. Consequently, a better understanding of and prediction capabilities for CHF in this regime is crucial for ensuring system safety and performance.

In this work, a new empirical correlation for predicting dryout-type CHF is developed based on an experimental CHF database covering R134a, CO₂, and water for reduced pressures of 0.7 to 0.99. Application of the correlation to a self-compiled collection of CHF data from the literature, which contains 2,880 data points for dryout in the near-critical pressure range, demonstrates excellent predictive capabilities. The prediction accuracy is evaluated using the mean relative error (MRE), mean absolute relative error (MARE), and root mean square error (RME), and compared against several existing correlations from the literature. The new correlation consistently outperforms the existing models across all metrics, achieving lower prediction errors for the intended range of application.

In CFD simulations, a consideration of buoyancy depends on a choice of density model. Quite often, the fluid is assumed incompressible and the Boussinesq model for density is used, where a fluid is assumed to have constant density while the buoyancy force is computed from the thermal expansion coefficient for a given temperature difference. Alternatively, fluid may be modelled with a temperature-dependent density which sometimes affects the stability of simulations, while for incompressible gases consideration of incompressible ideal gas law for computation of density is another alternative. In practice, the linear Boussinesq model is preferable choice, while its use is recommended only for fluids with small temperature differences. There is no universal criterium to verify whether the use of Boussinesq approximation is still appropriate or not.

This work is motivated by the analyses of an ex-vessel loss of coolant accident (LOCA) scenarios for the European Demonstration Fusion Power Plant (DEMO), where the consequences of postulated cooling loss of breeding blankets (BB) are mitigated with the injected gas which circulates inside the vacuum vessel (VV) transferring heat from the in-vessel components to the actively cooled VV. In LOCA scenario, fluid exhibits a large range of temperatures with temperature differences in order of few hundred degrees. Using the Boussinesq approximation in to large temperature range may overpredict the buoyancy force and the strength of natural convection, which is not conservative in respect to computed removed heat.

The objective of this study is to investigate how the chosen density model and the mesh resolution influence the CFD prediction of flow and heat transfer characteristics of buoyancy-driven flows. For this purpose, three simplified geometries, representing (i) the 2 cm gap between the breeding blanket and the vacuum vessel wall, (ii) free volume inside the VV and (iii) a simplified tokamak cross section geometry with BB segments inside, were analysed. The simulation results with temperature-dependent density model obtained by fitting of NIST database were compared with the simulation results with Boussinesq model, where different values of thermal expansion coefficients were tested. The above-mentioned aspects have been investigated with the steady-state CFD analysis.

Accurate evaluation of heat losses is a prerequisite for the correct interpretation of thermal-hydraulic experiments. When the fluid is colder than the ambient air, heat from the ambient air can be transferred to the fluid flow through the wall of the test section. The FEDORA experimental setup is designed to provide a high heating power through its heating elements, so the contribution of environmental heat gains is in general minor compared to the energy supplied by the heating system. The quantification of the heat gains is important for precise interpretation of thermal measurements. In this study, the environmental heat gains and the pressure drop of refrigerant R245fa, are numerically calculated, using a computational fluid dynamics code Ansys Fluent R1 2025, and compared with experimental measurements obtained in the FEDORA test section, constructed at the Reactor Engineering division R4 at Jozef Stefan Institute. The computational domain represents a simplified geometry in which the wall is modelled as a single homogeneous solid, omitting bolts, inserts, and internal support structures. Three meshes of increasing density are used to assess the influence of spatial resolution on the predicted heat transfer and pressure drop. The simulations are carried out at an inlet velocity of 3.02 m/s, an inlet temperature of 12.2 °C, and an outlet pressure of 2 bar, with the external wall temperature fixed at a measured value of 20.8 °C. The results are compared against experimental data to establish the accuracy of the simplified numerical approach in reproducing the small but measurable environmental heat input and pressure drop. The validity of the simulated and measured values has been checked through an analytical calculation method, placing the values between the purposefully exaggerated calculated heat fluxes. The outcomes of this work will be used in preparation of the numerical models, that are being developed at the Reactor Engineering division of the Jozef Stefan Insitute, needed to accurately simulate boiling in the fusion reactor divertor conditions.

Accurate modelling of droplet and bubble dynamics is essential for improving the predictive capabilities of condensation, boiling and other heat transfer simulations in nuclear thermal-hydraulics systems. Wettability, characterized by the contact angle, has a significant influence on heat transfer behaviour at solid–fluid interfaces. In realistic scenarios, dynamic wetting phenomena critically affect droplet or bubble nucleation and growth, with the evolution of the three-phase contact line during phase change process.

In this study, we present isothermal numerical simulations of single water droplet impacts on flat solid surfaces with varying wetting properties, using the OpenFOAM code. Special emphasis is placed on modelling the dynamic contact angle during the spreading and recoiling phases, as well as evaluating the effects of surface roughness and hydrophobicity on droplet behaviour. Key parameters, including maximum spreading diameter, contact time, and recoil dynamics, are quantified and compared to experimental measurements obtained with high-speed camera in THELMA laboratory, as well as with available data from the literature. The results provide valuable insights for the future development of more realistic models incorporating dynamic wettability effects at solid-fluid boundaries