Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) Project
Completed
Joint project

View of the ATLAS experimental loop. KAERI, Korea

The Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) is a thermal-hydraulic integral effect test facility for advanced light water reactors (LWRs) operated by the Korea Atomic Energy Research Institute (KAERI).

The ATLAS project series aims at investigations in the thermal-hydraulics 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/

ATLAS Members' area (password protected | reminder)

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

ATLAS-2 Members' area (password protected | reminder)

Third phase (2021-2024)

Tests to investigate design-basis accident (DBA) and beyond-design-basis accident (BDBA) transients have been conducted in the facility since 2014 as part of the NEA ATLAS project. For phase 3 of the project (2021-2024), the facility was upgraded with the addition of a scaled and instrumented containment mock-up, called CUBE, connected to the reactor coolant system (RCS).

Phase 3 of the project, involving 18 partners from 10 countries, was completed as planned at the end of 2024 and addressed knowledge gaps identified in previous ATLAS phases and in other NEA joint projects on nuclear thermal-hydraulic safety, e.g., the PKL and ETHARINUS Projects.

Tests included tests to characterise the CUBE facility performance (1 test), to investigate coupling between the RCS and containment for steam line break (SLB) and LOCA scenarios (2 tests), response of passive safety systems (PSS) for small break LOCA (SB-LOCA), intermediate break LOCA and SLB (3 tests), the effect of asymmetric natural circulation on cooldown (1 test) and designed extension conditions (DEC) for a SB-LOCA (1 test) and a total loss of heat sink (1 test).

Phase 3 extended the database on the effectiveness of passive heat removal safety systems; investigated reactor safety during design extension condition scenarios and asymmetric natural circulation phenomena; and provided data to address scaling as well as data on the coupling between the reactor coolant system (RCS) and containment responses during accident transients.

In addition, the JAEA provided as an in-kind contribution the test results of the large-scale test facility (LSTF) TR-LF-15 test performed in its ROSA LSTF. A counterpart test was performed in the ATLAS facility, as part of the ATLAS-3 project, allowing for scaling studies.

Benchmark and sensitivity crosswalk analyses were performed on one of the ATLAS-3 tests investigating the efficiency of PSS. The results of these analytical activities were jointly discussed with the ETHARINUS project at a joint analytical workshop held in Barcelona in November 2023. In 2024, a summary report of these activities was prepared for publication.

Discussions for a fourth project phase (2025-2028), with the objective of extending the integral tests database generated through previous phases, are well advanced. The new phase would include 10 integral tests, providing high fidelity data for thermal-hydraulic code enhancement. Topics would include the performance of containment spray systems, passive containment cooling system (PCCS) and in-containment refueling water storage tank (IRWST) safety injection through RCS – containment integrated tests, performance of PSS in SB-LOCA and station black-out (SBO), the effect of asymmetric natural circulation on cooldown, evaluations of accident management for complex DEC scenarios and scaling with the realisation of counterpart tests.

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

ATLAS-3 Members' area (password protected | reminder)

Participants

ATLAS: Belgium, China, Finland, France, Germany, Hungary, India, Japan, Korea, Russian Federation, Spain, Sweden, Switzerland, United Arab Emirates and the United States.

ATLAS-2 and ATLAS-3: Belgium, China, Czechia, France, Germany, Korea, Spain, Switzerland, the United Arab Emirates and the United States.

Project period

ATLAS: April 2014 - March 2017

ATLAS-2: October 2017 - September 2020

ATLAS-3: January 2021 - December 2024

Budget

ATLAS: EUR 2.5 million

ATLAS-2: EUR 3 million

ATLAS-3: EUR 4 million

Contacts