Myrrha primary heat exchanger tube rupture: Phenomenology and evolution

In the framework of MYRRHA Project, a pool-type experimental and material testing irradiation facility operated with Lead Bismuth Eutectic (LBE) coolant and able to operate in both sub-critical and critical mode is designed to be built in Mol, Belgium, in SCK\u2022CEN domain. In addition to the material testing function, targets of the MYRRHA reactor are to prove the feasibility of the ADS technology as Minor Actinides (MAs) burner and to act as a demonstrative plant for future Gen-IV heavy metal cooled reactors. SCK\u2022CEN entered the pre-licensing phase for the MYRRHA reactor. In order to provide the safety authority all the required data, a complete safety analysis must be performed, studying the transients defined by the list of postulating initiating events. In particular, an accident with potential serious consequences is the Primary Heat Exchanger Tube Rupture (PHXTR), involving the sudden release of single phase or two-phase water from a tube break in a hot liquid metal pool. This accident evolution is strongly characterized by the design of the MYRRHA Primary Heat eXchanger (PHX) and its direct surroundings in the reactor vessel and by the thermal-hydraulical conditions of the MYRRHA primary and secondary cooling system. In the first phase of a PHXTR accident, the water in the Secondary Cooling System (SCS) is released in the Primary System (PS) pool in regime of choked flow due to the pressure difference. Being the water released in an overheated, low-pressure environment, a flashing with potential sudden specific volume increase is expected. The heat transfer phenomena leading to the phase change velocity depend by the actual number of bubbles released in the hot liquid metal pool, function of the actual break size and shape. Its characterization is important for the definition of the overall specific volume increase and for the estimation of the water mass fraction redirected through the Primary Pump in the reactor Lower Plenum, with the risk of void insertion in the core and consequent reactivity excursion. A simplified calculation model to evaluate the history of any given bubble distribution generated by any water flow rate through any break has been set up. The main purpose is to describe the evolution of the main system state variables during the accidental event, by checking the potential insurgency of any reactor safety issue due to pressure peaks or core void insertions

Coupled system thermal Hydraulics/CFD models: General guidelines and applications to heavy liquid metals

This work aims to review the general guidelines to be adopted to perform coupled System Thermal Hydraulics (STH)/CFD calculations. The coupled analysis is often required when complex phenomena characterized by different characteristic time and length scales are investigated. Indeed, by STH/CFD coupling the main drawbacks of both stand-alone codes are overcome, reducing the computational cost and providing more realistic solutions. A review of several works available in literature and involving different coupling approaches, codes, time-advancing schemes and application fields is given. Besides STH/CFD coupling techniques, spatial domains and numerical schemes are analysed in detail. A brief description of applications to heavy liquid metal systems is also reported; lessons drawn in the frame of these and other works are then considered in order to develop a set of good practice guidelines for coupled STH/CFD applications

Calculation tool for heat exchanger tube rupture in pool-type reactors

In the framework of MYRRHA Project, a pool-type experimental and material testing irradiation facility operated with Lead Bismuth Eutectic (LBE) coolant and able to operate in both sub-critical and critical mode is designed to be built in Mol, Belgium, in SCK\u2022CEN domain. SCK\u2022CEN entered the pre-licensing phase for the MYRRHA reactor. A complete safety analysis must be performed to provide the necessary data to the safety authority. An event with potential serious consequences is the Primary Heat Exchanger Tube Rupture (PHXTR), involving the release of water from a failed tube in a hot liquid metal pool. In the first phase of a PHXTR accident, the water in the Secondary Cooling System is released in the Primary System pool in choked flow regime. Afterwards, the water bubble formation and characterization is important for the definition of the water specific volume increase and for the estimation of the water mass fraction redirected in the reactor Lower Plenum, with the risk of void insertion in the core and consequent reactivity excursion. An analytical calculation model to evaluate the evolution of any bubble distribution has been set up. The main purpose is to describe the evolution of the main system variables during the accidental event, by checking the potential insurgency of any reactor safety issue due to pressure peaks or core void insertion

H2020 MYRTE CIRCE-HERO experimental campaign: post-test activity and code validation

In the frame of H2020-MYRTE (MYRRHA Research and Transmutation Endeavour) project, the LBE CIRCE facility, at ENEA-Brasimone research center, has been reconfigured with the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, consisting in a seven double wall bayonet tube bundle heated by LBE on the shell side and by water on the tube side. Despite having been conceived for tests aimed to the assessment of the ALFRED reactor heat exchanger, the experimental campaign in the MYRTE project foresees the facility to be used to simulate the behavior of the MYRRHA Primary Heat Exchanger in its innovative double-walled bayonet tube configuration. The thermal-hydraulic parameters characterizing the system are modified to represent MYRRHA plant conditions in terms of temperatures, pressures, void fraction and flow regimes. Its main purpose consists in studying the heat transfer process in the bayonet tube to characterize the convective and the conductive heat transfer processes and in comparing the experimental data with calculation models, for validation purposes. A RELAP5-3D model of the CIRCE-HERO facility has been realized and then used for a series of steady state pre-tests, which have been then compared with the experimental results performed in the MYRTE campaign. A number of discrepancies have been identified and explained. Finally, the model has been updated through the experimental feedback and the simulation accuracy has notably increased in terms of energy and momentum balance; the RELAP5-3D model is then able to provide an accurate experimental data representation

Shielding and activation calculations around the reactor core for the MYRRHA ADS design

In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions

Shielding and activation calculations around the reactor core for the MYRRHA ADS design

In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions

Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF

The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental-scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facilities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall system that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Several parameters were analyzed, like the criticality behavior (namely the K-eff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed

MYRRHA primary heat exchanger experimental simulations on CIRCE-HERO

In the frame of the MYRRHA ADS technological development, an intense research effort has been undertaken by European Commission, in particular for the Heavy Liquid Metal technology assessment. In this scenario, EURATOM HORIZON2020 funded the SESAME & MYRTE projects, coordinating a series of thermal hydraulics experiments and simulations for the safety assessment of liquid metal fast reactors. In particular, in the MYRTE project, a dedicated activity has been defined,consisting of a low pressure experimental campaign realized at the ENEA Brasimone Research Centre and concerning a double wall Steam Generator Bayonet Tube (SGBT), aiming at supporting the development of the Primary Heat Exchanger (PHX) of MYRRHA.The CIRCE (CIRColazione Eutettico) pool facility hosts a devoted Test Section (TS) named HERO (Heavy liquid mEtal pRessurized water cOoled tubes). HERO consists of seven double wall bayonet tubes, with an active length of 6 m, arranged in a hexagonal geometry. This solution aims at improving the plant safety reducing the possibility of water-lead/lead alloy interaction thanks to a double physical separation between them and allowing an easier control of potential leakages of the coolant by pressurizing the separation region with inert gas.The experimental campaign in CIRCE-HERO has been supported by a pre-test analysis using system thermal-hydraulic codes (i.e. RELAP5). A sensitivity analysis has been performed considering the mass flow rate and HERO inlet temperature for both the water and LBE, fixing the operating pressure at 16 bar.A test matrix, as produced as output of the sensitivity analysis and aimed at reproducing as faithfully as possible the thermal-hydraulic behaviour of the MYRRHA PHX under different operating conditions, is here presented. The experimental results collected from the low pressure tests on HERO test section are reported and discussed