Modeling of Table Top Primary Heat Exchanger Experiment RUN#5 by Means of the System Codes RELAP5 mod 3.3 and Flownex SE 8.4

Davide Rozzia, Marinus Potgieter, Jan Cools, Tom Van Loy, Graham Kennedy, Julio Pacio, Katrien Van Tichelen, Rafaël Fernandez

Research outputpeer-review

Abstract

MYRRHA is a flexible, fast-spectrum, pool-type research reactor cooled by lead-bismuth eutectic (LBE), under development at SCK CEN. The Primary Heat eXchangers (PHX) of MYRRHA are designed to operate with pressurized water as secondary coolant and are located in the reactor pool. Reducing the probability of LBE-water interaction requires the PHX design to be based on a doublewalled bayonet tube bundle, with a Helium leak detection system. This specific configuration has never been tested in the nuclear technology sector and therefore needs experimental validation in order to cover the R&D gap. For this purpose, SCK CEN has set up an experimental program with different facilities dedicated to assess the PHX performance. The so-called Table Top facility is a 2.5 kW water/steam open loop under operation at SCK CEN and works with natural circulation flow in the pressure range [8 -16] bara. The objective of the facility is to obtain a first estimation of the thermal resistance of the PHX double-walled tube and provide data for code assessment and model development. During the commissioning phase of the installation, six different runs where executed to investigate the water/steam flow in the operational pressure range. Among these tests, RUN#5 demonstrated the capability of the facility to operate with stable flow at 16 bara at its maximum power. The paper provides the results of test RUN#5 and investigates the capability of system thermal-hydraulic codes to represent the inception and development of two-phase flow in natural circulation. Two codes used at SCK CEN to design experimental facilities as well as MYRRHA secondary cooling system are benchmarked against RUN#5 test data: RELAP5 mod 3.3 and Flownex SE 8.14. The choice to use two codes is also made in the interest of experimental validation and R&D qualification.
Original languageEnglish
Title of host publication20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20)
PublisherAmerican Nuclear Society
Pages722-735
Number of pages14
DOIs
StatePublished - 18 Jan 2024

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