Conference Agenda

SMRs and advanced reactors concepts can involve specific design characteristics requiring modelling capabilities that are beyond the validated boundaries of existing codes or include phenomena for which the existing experimental data is insufficient. The significant efforts and resources associated with performing validation or experimentation constitutes a challenge to a safe and secure timely deployment of SMRs. To overcome those challenges, the Internation Atomic Energy Agency (IAEA) set up a working group within the Nuclear Harmonization and Standardization Initiative (NHSI) to establish a Network for Experiments and Code Validation for Design and Safety Analysis Computer Codes for SMR and Advanced Reactor Designs (NEXSHARE). NEXSHARE is a technical forum of global cooperation and resource sharing for experiments and code validation between entities operating experimental facilities, design organizations of SMRs, Regulators’ Technical Support Organizations (TSOs) and other International Organizations. In particular, OECD Nuclear Energy Agency (NEA) and the Generation IV International Forum (GIF) are closely collaborating to this project. NEXSHARE was launched in 2024 at the IAEA Workshop on Experimental Testing and Validation for Design and Safety Analysis Computer Codes for SMRs. Feedback from the participants helped shape the next steps for the Network which include optimizing its functionalities, expanding its experimental facilities database, and conducting technology specific efforts on experiments and code validation. This paper provides an overview of NEXSHARE’s design and functionalities and also outlines the Network usages and benefits for the industry, supporting the IAEA’s initiatives to accelerate the development and deployment of safe and secure advanced reactors, including SMRs.

The McSAFER project was focused on experimental and numerical investigations for the safety evaluation of water-cooled SMRs such as NuScale, SMART, CAREM and F-SMR. The experimental program consisted in test series at three EU facilities e.g., MOTEL at LUT, HWAT at KTH and COSMOS-H at KIT. The experimental data was used for the validation of thermal hydraulic codes (CFD, subchannel and system codes). The experiments covered safety-relevant phenomena such as cross-flow in the core, the performance of the helical-coiled heat exchanger, forced and natural circulation and its transition, etc. The numerical part was devoted to the analysis of the core behavior under normal and accidental conditions (REA, Cold water injection) of four core designs (CAREM, NuScale, KSMR and F-SMR) using both industry-like and advanced transport Multiphysics computational routes. The behavior of a NuScale core loaded with ATF fuel under REA-conditions was investigated with three different high-fidelity coupling of neutronic, thermo-mechanics and thermal hydraulic codes and the obtained results were compared to the ones predicted for a core loaded with UO₂. Finally, selected transients (Steam Line Break for NuScale and SMART) were analyzed with three multiscale / multiphysics coupled codes including system TH, subchannel TH, CFD and 3D nodal diffusion codes. This paper will present and discuss the main outcomes of the core and plant analysis emphasizimg the capabilities and future improvements for a more realistic prediction of safety parameters of SMRs as well the potentials of the methods for the analysis of transients in nuclear power plants of Gen-2 and -3.

In the design and development of i-SMR, the passive emergency core cooling system (PECCS) is quite different from that of the existing reactors, and in particular, the depressurization valves and the recirculation valves may have completely different configurations and components from the existing ECCS valves.

The reason for such a complex configuration is that not only should the valves be able to be opened passively, but also actively opened by the actuation signals, and undesired opening should be prevented even with a single failure of the component.

Dynamic behavior of main valve consisting of spool discs, springs and orifices, block valve of specific shape, actuator trip valve and connecting pipes, etc., is critical at the validation of the design.

In this study, for this problem, the pressure and flow rate at each flowing part of the valve were calculated using the MARS-KS code, and the equations of motion of the spool disks were solved using the calculated flow data to determine the opening area of each valve, and the dynamic behavior was analyzed over time by feeding it back to the MARS-KS code calculation. The scheme was named as SEMICOM (Solution of Equation of Motion Implemented by Control-variables Of MARS code).

Using this method, it analyzes the dynamic response of a virtual PECCS valve and provides a requested performance data that can help determine various design parameters such as spring constant, disk-cylinder gap, and orifice size as well as dynamic stability determination.

The Horizon Euratom EASI-SMR project (Ensuring Assessment of Safety Innovations for light water SMR) aims to address critical R&D needs for the safety demonstration of Light Water (LW) Small Modular Reactor (SMR) technology, supporting its short-term deployment in Europe. Focusing on the European designs NUWARD and LDR-50, EASI-SMR targets innovations such as passive safety systems, boron-free cores, co-generation, additive manufacturing, and multi-unit operation. The project’s goal is to ensure that these reactors are designed, constructed, and licensed in accordance with European regulatory standards.

This paper discusses the core of the EASI-SMR project, which consists of three interconnected work packages: WP2 – Experimental Testing Program, WP3 – Code Validation and Scaling, and WP4 – Reliability of Passive Systems. WP2 establishes a new experimental program to investigate key physical phenomena in passive safety systems under both design basis and beyond design basis conditions, providing essential insights for LW-SMR safety demonstration. In WP3, the capability of European-developed codes to simulate DBA and BDBA scenarios is assessed, alongside the identification of best practices for passive system modeling and areas for code development. Finally, WP4 applies these validated codes to perform reliability assessments, focusing on risk analysis and licensing readiness for passive systems. This structured process, from experimentation in WP2 to code validation in WP3 and reliability assessment in WP4, creates a comprehensive and interconnected framework that addresses R&D needs, supporting the short-term deployment of LW-SMRs across Europe.

Most of the previously proposed data-driven fault diagnosis methods are Close Set Recognition (CSR) methods, which assumes that the training set and test set are drawn from the same fault label space. The resulting problem is that when facing an unknown type of fault that not included in the training set, CSR method will incorrectly classify it as one of the known fault types in the training set, bringing a huge negative impact on actual fault diagnosis tasks of nuclear power plants. The fault types of nuclear power plants cannot be exhaustive, and the fault types included in the training set are limited due to the difficulties in collecting and labelling data. Therefore, almost all nuclear power plant fault diagnosis tasks are essentially Open Set Recognition (OSR) tasks, which requires not only the correct classification of known fault types, but also the identification of unknown fault types. However, there are few related researches on OSR fault diagnosis in nuclear power plants.

To solve the above dilemma, a novel nuclear power plant OSR fault diagnosis framework based on CPL is proposed. Experimental data of 10 health states and 841 monitoring variables are generated by a detailed digital nuclear power plant model, which can truly reflect the high dimensionality and strong nonlinearity characteristics of nuclear power plant data. And 24 OSR tasks with different settings of known and unknown fault types are designed, on which the feasibility and effectiveness of the proposed framework are verified.

Heat pipe cooled reactors have garnered significant attention due to their simple design, scalability, and reliability, making them an ideal choice for nuclear power generation in space and deep-sea applications. The integration of a supercritical CO2 Brayton cycle system with heat pipes meets the demand for system miniaturization and high energy conversion efficiency in nuclear power systems. Although several conceptual designs have been proposed based on this idea, there is still a lack of research on the operational characteristics of these reactors, particularly concerning the impact of high-temperature heat pipe oscillations on system performance.

In this study, a coupled code combining a heat pipe-cooled reactor and a Brayton cycle system was developed to assess the transient effects of heat pipe temperature oscillations on system performance. The reactor code includes a neutron physics model, a heat pipe model, and a core heat transfer model, which were validated using reference data and experimental results. The supercritical carbon dioxide Brayton cycle system was modeled using a customized version of the Relap5 code, and the coupling between the two subsystems was successfully implemented.

Simulation results reveal that heat pipe temperature oscillations induce synchronous oscillations in the reactor and Brayton cycle system’s operational parameters, such as temperature and pressure. However, the operational state of the Brayton cycle system is less affected compared to that of the reactor system. This coupled code serves as an effective tool for the design and safety analysis of supercritical CO2 heat pipe cooled reactors.