Prospective For Nuclear Thermal Hydraulic Created By Ongoing And New Networks

International audienceThis paper introduces the FONESYS, SILENCE and CONUSAF projects run by some of the leading organizations working in the nuclear sector.The FONESYS members are developers of some of the major System Thermal-Hydraulic (SYS-TH) codes adopted worldwide, whereas the SILENCE members own and operate important thermal-hydraulic experimental facilities. The two networks work in a cooperative manner and have at least one meeting per year where top-level experts in the areas of thermal-hydraulic code development and experimentation are gathered.The FONESYS members address various topics such as hyperbolicity and numerics in SYS-TH codes, 3-field modeling, transport of interfacial area, 3D modeling, scaling of thermal-hydraulic phenomena, two-phase critical flow (TPCF), critical heat flux (CHF), and others. As part of the working modalities, some numerical benchmarks were proposed and successfully conducted by the network, addressing some of the most relevant topics selected by the FONESYS members.On the other hand, SILENCE addresses topics such as identification of current measurement needs and main gaps for further SYS-TH and CFD codes development and validation, definition of similar tests and counterpart tests in Integral Tests Facilities (including containment thermal-hydraulics) to be possibly conducted on Members' test facilities, scaling issue, and other subjects. Furthermore, SILENCE organized a Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics (SWINTH) which was held in Italy on June 2016. A second edition of the Workshop, namely SWINTH-2019, will be held in Italy in 2019 under the umbrella of the OECD/NEA/CSNI/WGAMA.Recently a new initiative is being taken by launching an international consortium of nuclear thermal-hydraulics code users, the CONUSAF. The main idea is to enhance the interactions between the users of computational tools in nuclear TH, noticeably including SYS-TH and CFD codes, the code developers and the experimentalists. The proposed initiative is expected to have a positive impact on the entire ecosystem by pursuing the assessment of the current code limitations and capabilities, analyzing and addressing issues raised by the users and promoting common RandD efforts on topics of high relevance

PARTICLE DEPOSITION STUDY IN A HORIZONTAL SQUARE CHANNEL USING HIGH-RESOLUTION IMAGING

The ongoing research at Texas A&M University (TAMU) aims to develop and implement measurement techniques to quantify fission product transport and related phenomena in a Gas-Cooled Fast Reactor (GCFR). This study summarizes the experimental measurements performed in a Proof-of-Concept Facility (PCF) to understand particle transport under various Reynolds numbers. Using high-resolution high-magnification optical measurement techniques, surrogate particle deposition was measured on the bottom wall surface of the PCF. The particle deposition over time was observed to decrease as the Reynolds number increased. In addition, the particle spatial distributions increased approaching the side wall of the PCF. This state-of-the art optical technique allows for the non-intrusive measurements of micron-sized particles near the wall surface

PARTICLE DEPOSITION STUDY IN A HORIZONTAL SQUARE CHANNEL USING HIGH-RESOLUTION IMAGING

The ongoing research at Texas A&M University (TAMU) aims to develop and implement measurement techniques to quantify fission product transport and related phenomena in a Gas-Cooled Fast Reactor (GCFR). This study summarizes the experimental measurements performed in a Proof-of-Concept Facility (PCF) to understand particle transport under various Reynolds numbers. Using high-resolution high-magnification optical measurement techniques, surrogate particle deposition was measured on the bottom wall surface of the PCF. The particle deposition over time was observed to decrease as the Reynolds number increased. In addition, the particle spatial distributions increased approaching the side wall of the PCF. This state-of-the art optical technique allows for the non-intrusive measurements of micron-sized particles near the wall surface

Prospective for nuclear thermal hydraulics created by ongoing and new Networks

The purpose of the present article is to introduce the FONESYS, SILENCE and CONUSAF projects run by some of the leading organizations working in the nuclear sector. The FONESYS members are developers of some of the major System Thermal-Hydraulic (SYS-TH) codes adopted worldwide, whereas the SILENCE members own and operate important thermal-hydraulic experimental facilities. The two networks work in a cooperative manner and have at least one meeting per year where experts of the highest level in the areas of thermal-hydraulic codes development and experimentation are gathered. The FONESYS members address various topics such as hyperbolicity and numerics in SYS-TH codes, 3-field modeling, transport of interfacial area, 3D modeling, scaling of thermal-hydraulic phenomena, two-phase critical flow (TPCF), critical heat flux (CHF), and others. As part of the working modalities, some numerical benchmarks were proposed and successfully conducted by the network addressing some of the most relevant topics selected by the FONESYS members (i.e. code developers). On the other hand, SILENCE addresses topics such as current measurement needs and main gaps for further SYS-TH and CFD codes development and validation, working on the identification of similar and counterpart tests in Integral Tests Facilities (including containment thermal-hydraulics) to be possibly conducted on Members’ test facilities, but also dealing with the scaling issue and other subjects. Furthermore, SILENCE organized a “Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics (SWINTH)” which was held in Italy on June 2016. A second edition of the Workshop, namely SWINTH-2019, will be held in Italy in 2019 under the umbrella of the OECD/NEA/CSNI/WGAMA. Recently a new initiative is being taken by launching an international consortium of nuclear thermal-hydraulics code users, the CONUSAF. The main idea is to enhance the interactions between the users of computational tools in nuclear TH, noticeably including SYS-TH and CFD codes, the code developers and the experimentalists. The proposed initiative is expected to make a positive impact on the entire ecosystem by pursuing the characterization of the current code limitations and capabilities, analyzing and addressing issues raised by the users and promoting common R&D efforts on topics of high relevance

Iaea's coordinated research projects on thermal hydraulics of fast reactors

A Coordinated Research Project on “Benchmark Analysis of FFTF Loss of Flow Without Scram Test” was launched by the International Atomic Energy Agency (IAEA) in 2018. A series of passive safety tests were conducted from 1980-1992 at the Fast Flux Test Facility (FFTF), 400 MW(th) liquid sodium cooled nuclear test reactor owned by U.S. Department of Energy (DOE) to demonstrate the potential of FFTF to survive severe accident initiators with no core damage. Amongst these tests was a series of Loss of Flow Without Scram (LOFWOS) tests from power levels up to 50%, also commonly referred to as Unprotected Loss of Flow (ULOF) tests, which were studied in the IAEA CRP. The data were provided by the Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL). Another Research Coordinated Project on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” was launched by the IAEA in 2022. Three tests were conducted in 2017 to study the thermal-hydraulic behavior of a test fuel assembly cooled by lead-bismuth eutectic alloy during transition from forced to natural convection at the NACIE-UP facility at Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy. This project is the first IAEA CRP that is dedicated to the thermal hydraulics of lead and lead bismuth eutectic (LBE) technology. The paper provides a general overview of the two CRPs within the framework of the IAEA activities on thermal hydraulics of fast reactors