^{View of the ATLAS experimental loop. KAERI, Korea}

The Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) Project was referenced in The Fukushima Daiichi Nuclear Power Plant Accident: OECD/NEA Nuclear Safety Response and Lessons Learnt report as one of the three new joint projects based on existing research facilities which address safety issues related to the Fukushima Daiichi accident.

The ATLAS project is aimed at investigations in the thermalhydraulics field of accident scenarios of high safety relevance for both existing and future nuclear power plants.

Such investigations are of specific importance because they contribute to the validation of computer codes that are required in safety evaluation of light water reactors (LWRs) in order to simulate plant behaviour during design basis accidents (DBAs) and design extension conditions (DECs). They cover complex multi-dimensional single-phase and two-phase flow conditions.

Although current thermal-hydraulic safety analysis codes have achieved very high predictive capability especially for one-dimensional phenomena, there is a strong need for experimental work and code development and validation for more complex flow conditions.

Moreover, the increased use of best-estimate (BE) analysis methods in licensing, which is replacing traditional conservative evaluation model (EM) approaches, require the validation and quantification of uncertainties in the simulation models and methods.

Many experimental facilities have contributed to the thermal-hydraulic databases available today which have been extensively used for the validation of EM and BE computer codes. However, most of the current data are insufficient for future codes which aim at multi-dimensional simulation capabilities, mainly because the spatial resolution of measurement is not sufficient to assess the simulation models and methods, especially for integral system testing.

ATLAS phases

First phase (2014-2017)

The main objective of the first phase was to provide experimental data for resolving key LWR thermal-hydraulics safety issues related to multiple high-risk failures and highlighted in particular from the Fukushima Daiichi nuclear power plant accident, by using the ATLAS facility at the Korea Atomic Energy Research Institute (KAERI).

The first phase focused in particular on the validation of simulation models and methods for complex phenomena of high safety relevance to thermal-hydraulic transients in DBAs and DECs, more particularly:

• it generated an integral system and separate-effect experimental database to validate the predictive capability and accuracy of computer codes and models. Thermal-hydraulic phenomena coupled with multi-dimensional flows that included mixing, stratification, counter-current flows, parallel-channel flows and oscillatory flows were the main focus of the investigations;
• it contributed to the assessment of codes in use for thermal-hydraulic safety analyses, as well as advanced codes under development, including three-dimensional computer codes, through active involvement of the project partners, who maintained and improved the technical competence in thermal-hydraulics for nuclear reactor safety (NRS) evaluations.

The experimental programme provided a valuable and broadly usable database to achieve the above objectives. In phase one, a total of eight tests at the ATLAS facility were performed within five different research topics:

• prolonged station blackout (SBO)
• small break loss-of-coolant accident (SBLOCA) during SBO
• total loss of feedwater (TLOFW)
• medium-break LOCAs
• scale-up issues – related to assessing the applicability of small-scale experimental data to full-scale reactors.

The experimental programme and associated analytical activities helped creating an analytical group among NEA member countries which share the need to maintain or improve the technical competence in thermal hydraulics for nuclear reactor safety evaluations. Within the group activities, a joint workshop with the OECD/NEA PKL3 project was organised which main outcomes are presented here.

The summary report of phase 1 is available here.

The Data package for phase 1 is available upon request to the NEA Databank at https://www.oecd-nea.org/tools/abstract/detail/csni2039/.

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Second phase (2017-2020)

ATLAS-2 was a follow-up to phase one and was focused on topics of high safety relevance for both existing and future nuclear power plants that had been identified by the participants. The following topics were addressed:

• long-term coolability with partial core blockage
• passive core makeup during station blackout (SBO) and small break loss of coolant accident (SBLOCA)
• intermediate break loss of coolant accident (IBLOCA), including risk-informed break size definition
• design extension condition (DEC) scenarios such as steam line break (SLB) followed by steam generator tube rupture (SGTR) and shutdown coolability without residual heat removal system (RHRS)
• open test to address scaling issues by performing counterpart test to previous Integral Effects Tests (IETs).

The experimental programme was designed to provide an integral-effect experimental database to support the enhancement of code predictive capability and accuracy of models. In phase 2, a total of 8 tests were performed. Analytical activities were pursued including the oganisation of a joint workshop with the OECD/NEA PKL4 project which outcomes are provided here.

The summary report of phase 2 is available here.

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Third phase (2021-2024)

ATLAS-3 is a follow-up to phase two and aims to enhance the nuclear safety analysis methods and improve the best guidelines for accident management. Investigations are focusing on remaining key issues that have been identified following the completion of previous phases. ATLAS-3 focus on the validation of simulation models and methods for addressing complex phenomena of high safety relevance to thermal-hydraulic transients in DBA and BDBA. A new feature of the project is that the facility has been extended with a containment mock-up system (CUBE) and offers the possibility to perform tests covering thermalhydraulics in the reactor coolant system and in the containment.

The experimental programme is intended to provide a valuable and broadly usable database. In the NEA ATLAS Phase 3 Project, a total of 10 tests by utilising the extended ATLAS facility are conducted in five different research topics:

• Integral Effect Tests on Interaction between Reactor Coolant System (RCS) and Containment, for Steam Line Break (SLB) and Intermediate Break Loss of Coolant Accident (IBLOCA);
• Passive Safety Systems during Small Break Loss of Coolant Accident (SBLOCA), Intermediate Break Loss of Coolant Accident (IBLOCA), and Steam Line Break (SLB);
• Natural Circulation, focusing on asymmetric cool-down condition;
• Design Extension Condition (DEC) scenarios such as Multiple Failure of Small Break Loss of Coolant Accident (SBLOCA) under Station Blackout (SBO) condition, and Total Loss of Heat Sink;
• Open test to address scaling issues by performing counterpart test to previous Integral Effects Tests (IETs).

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Fourth phase (2025-2028)

The ATLAS Phase-4 project seeks to advance international understanding of complex accident scenarios in water-cooled reactors, including small modular reactors (SMRs), with particular emphasis on the performance of passive safety systems within both the reactor coolant system (RCS) and the containment. A key objective is to enhance the validation of thermal-hydraulic simulation codes through a series of large-scale Integral Effect Tests (IETs) conducted at the ATLAS-CUBE facility which simulates both RCS and containment components.

Running from January 2025 to December 2028, the four-year programme of phase 4 of the ATLAS project will include ten IETs. Building on experience from previous ATLAS phases and shaped by post-Fukushima Daiichi priorities, particularly regarding passive system behaviour, ATLAS-4 targets five thematic research areas:

• interaction between the reactor coolant system and containment;
• performance evaluation of passive safety systems in the RCS and containment;
• effects and dynamics of natural circulation;
• assessment of design extension conditions (DECs);
• counterpart testing to support simulation code validation and benchmarking.

These tests aim to address persistent challenges such as low driving forces, asymmetric flow patterns, thermal-hydraulic instabilities, and the realistic functioning of passive systems under accident conditions.

By bringing together regulators, research institutions, and industry experts, ATLAS-4 will significantly deepen collective knowledge of safety-critical thermal-hydraulic phenomena and contribute to the advancement and credibility of safety analysis codes for both current and next-generation nuclear systems.

In parallel with the experimental programme, both blind and open benchmark exercises will be organised to evaluate different modelling approaches and quantify associated uncertainties. These activities aim to strengthen the predictive capabilities of simulation tools used in safety analyses.

To support knowledge dissemination, a series of analytical workshops will be held in collaboration with other NEA thermal-hydraulics projects – building on the successful experience of past initiatives such as joint workshops with the PKL [Primärkreislauf-Versuchsanlage (Primary Coolant Loop Test Facility)] project and the Experimental Thermal Hydraulics for Analysis, Research and Innovations in Nuclear Safety (ETHARINUS) project.

A strong emphasis will also be placed on education, training, and knowledge transfer to younger generations. A dedicated programme, to be developed under the Nuclear Education, Skills and Technology (NEST) framework, will support this objective. The NEST ATLAS initiative will foster international collaboration, capacity building, and professional development for young researchers and students. Through hands-on involvement in experimental and simulation-based training, as well as participation in analytical workshops, NEST participants will gain valuable experience and deepen their understanding of thermal-hydraulic safety research.

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Related research on thermalhydraulics projects can be found at Integral Test Facilities and Thermal-Hydraulic System Codes in Nuclear Safety Analysis.