Mariano Tarantino

Pool temperature stratification analysis in CIRCE-ICE facility with RELAP5-3D© model and comparison with experimental tests

In the frame of heavy liquid metal (HLM) technology development, CIRCE pool facility at ENEA/Brasimone Research Center was updated by installing ICE (Integral Circulation Experiments) test section which simulates the thermal behavior of a primary system in a HLM cooled pool reactor. The experimental campaign led to the characterization of mixed convection and thermal stratification in a HLM pool in safety relevant conditions and to the distribution of experimental data for the validation of CFD and system codes. For this purpose, several thermocouples were installed into the pool using 4 vertical supports in different circumferential position for a total of 119 thermocouples [1][2].The aim of this work is to investigate the capability of the system code RELAP5-3D© to simulate mixed convection and thermal stratification phenomena in a HLM pool in steady state conditions by comparing code results with experimental data. The pool has been simulated by a 3D component divided into 1728 volumes, 119 of which are centered in the exact position of the thermocouples. Three dimensional model of the pool is completed with a mono-dimensional nodalization of the primary main flow path. The results obtained by code simulations are compared with a steady state condition carried out in the experimental campaign. Results of axial, radial and azimuthal temperature profile into the pool are in agreement with the available experimental data Furthermore the code is able to well simulate operating conditions into the main flow path of the test section.

Natural and Gas Enhanced Circulation Tests in the NACIE Heavy Liquid Metal Loop

The paper reports on the results carried out from the natural circulation and gas-injection enhanced circulation tests performed on a heavy liquid metal loop, named NACIE, and located by the Brasimone ENEA Research Centre. The work is aimed at providing information on the characterization and interpretation of the basic mechanisms proposed in the design of future reactor relying on these circulation mechanisms. The results discussed in the present work concern the experiments performed using Lead Bismuth Eutectic (LBE) as coolant. Both natural circulation and gas-injection enhanced have been addressed, drawing conclusions about the observed phenomena. Numerical simulations have been performed in collaboration with the University of Pisa, adopting the RELAP5/Mod3.3 system code modified to allow for LBE as a cooling fluid. Post-test calculations have been performed to compare the code response with the experimental results under the natural circulation and gas enhanced circulation flow regime, allowing to qualify the adopted nodalisation as well as the performance of the code when employed on HLM loop. The available data will allow to validate and qualify numerical tools for engineering application, establishing a reference experiment for the benchmark of commercial codes when employed in HLM loop.Copyright

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.

Experimental activity for the investigation of mixing and thermal stratification phenomena in the circe pool facility

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development.In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development.In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development.The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermomechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap.In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5].A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported.The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat.Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6–7].Copyright

Experimental Study on Natural Circulation and Air-Injection Enhanced Circulation With Different Fluids

This paper reports on an experimental investigation on natural circulation and air-injection enhanced circulation performed adopting different fluids. This work is aimed at providing information on the basic mechanisms proposed in the design of future reactors relying on such circulation mechanisms for core cooling. Though the final objective of the research is the study of heavy metal cooling, the work is here limited to nonmetallic fluids. The working fluid adopted in past analyses was water. Further experimental campaigns were recently performed using the Novec™ HFE-7100 fluid, providing additional information on basic phenomena and the related scaling laws. The new fluid has a greater density and a greater thermal expansion coefficient with respect to water. Air was adopted for gas injection. Both natural circulation and gas-injection enhanced circulation are addressed in this work, drawing quantitative conclusions about the observed parametric trends. A systematic comparison is performed with the results obtained in previous experimental activities using water.

Test Section Design for SGTR Experimental Investigation in CIRCE Facility for HLMRS Supported by SIMMER-III Code

In the framework of MAXSIMA project, the design of a large-scale Test Section (TS), aiming to experimentally investigate the Steam Generator Tube Rupture (SGTR) postulated event in a relevant configuration for Gen IV MYRRHA reactor, was carried out. The TS will be implemented in the large pool CIRCE facility, at ENEA CR Brasimone. The TS is composed of four tube bundles representing a full scale portion of the Primary Heat eXchanger (PHX) of MYRRHA plant. They allow the execution of four SGTR tests, one at a time, excluding the necessity to extract the TS from the facility after each test. Water is foreseen to be injected at 16 bar and 200°C in the pool, partially filled by LBE at 350°C with a cover gas of argon at about 1 bar.The pressurization transients of CIRCE vessel and the sizing of the discharge lines and relative rupture disks were numerically predicted by SIMMER-III code on the base of a preliminary simplified configuration of the TS. The obtained results showed that the design pressure of CIRCE main vessel was not reached during more than 10 s of water injection, implementing a singular rupture disk having a diameter of 2 inch activated at 6.5 bar. It appears more than enough to notice, in a real reactor, the occurrence of the SGTR event and stop the water supply, interrupting the accidental scenario. These numerical results were adopted to support the design of the presented TS.Copyright

Coupled simulations of natural and forced circulation tests in Nacie facility using relap5 and Ansys fluent codes

In this work the activity performed at the DICI (Dipartimento di Ingegneria Civile e Industriale) of the Pisa University in collaboration with the ENEA Brasimone Research Centre is presented. In particular the document deals with the application of an in-house developed coupling methodology between a modified version of RELAP5/Mod3.3 and Fluent commercial CFD code, to the NACIE (Natural Circulation Experiment) LBE experimental loop (built and located at the ENEA Brasimone research centre).The first part of the document treats the description of the NACIE loop type facility, while in the second part, the developed coupling tool is presented and the obtained numerical results are compared to stand alone RELAP5 results and to data obtained from the NACIE experimental campaign. The experimental tests are performed varying the argon flow rate and the electric power supplied to the heater and both natural and assisted circulation tests are investigated. The numerical model set-up is based on a two-way explicit coupling scheme and 2D and 3D geometrical domain were investigated.Comparative analyses among numerical and experimental results showed good agreement, giving positive feedback on the feasibility and capability of the developed coupling methodology.Copyright

A CFD Analysis of Flow Blockage Phenomena in ALFRED LFR DEMO Fuel Assembly

A CFD study has been carried out on fluid flow and heat transfer in the HLM-cooled Fuel Pin Bundle of the ALFRED LFR DEMO.In the context of GEN-IV Heavy Liquid Metal-cooled reactors safety studies, the flow blockage in a Fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR Fuel Assembly. The present paper is a first step towards a detailed analysis of such phenomena, and a CFD model and approach is presented to have a detailed thermo-fluid dynamic picture in the case of blockage. The closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED has been modeled and computed. At this stage, the details of the spacer grids have not been included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, have not been included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900°C, are reached with blockage larger than 30% in terms of area fraction.Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages ( 15% could be detected by putting some thermocouples in the plenum region of the FA.© 2014 ASME

HELENA: A Heavy Liquid Metal Multi-Purpose Loop for Thermal-Hydraulics, Corrosion and Component Test

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection. In this frame, the HELENA facility is well instrumented and it represents a loop working in pure lead for experiments in the field of corrosion for LFR structural materials, component test, and thermal-hydraulic investigations. The components and the process of the facility has been also depicted in some details. The scheduled working activities has been described and the main steps focused.A prototypical mechanical pump has been designed and manufactured to properly work in pure lead at high temperatures. The long-run test on the component will be isothermal at 400°C with low oxygen content <10−8 % in weight; the oxygen content will be monitored continuously. The pump will be gradually driven to the reference mass flow rate 35 kg/s and this mass flow rate will be maintained for 1500 h, i.e. 2 months about. After this, the pump is stopped and the loop is drained. Then, the pump impeller is disassembled by the body and it is analyzed for corrosion. Then, a test on the ball valves is carried out for a few months. At the end of 2014, the facility is upgraded with the insertion of the FPS in the heating section and with the secondary side. A 19-pin wire-spaced Fuel Pin bundle Simulator (FPS) is installed to measure clad temperature and heat transfer coefficients in different conditions in the different ranks of sub-channels of the MYRRHA bundle. Therefore a test matrix on the forced convection condition in the wire-spaced bundle will be carried out.A 7 tube/shell-and-tube Heat Exchanger couples the primary lead loop with the secondary side with water in pressure at 100 bar. The tube-in-tube technology with lead tube side, water shell side, steel powder in the gap is adopted. Bubble tubes with flowing Argon are adopted to measure pressure losses in the different branches of the loop. Several thermocouples monitor the loop in different points. An ancillary gas system ensures the cover gas.The paper reports the description of the experiments, the proposed test matrix and description the technological solution adopted for the HELENA implementation.Copyright

HLM Fuel Pin Bundle Characterization in CIRCE Pool Facility

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection, heat transfer and removal, component development and testing, remote maintenance, procedure definition and coolant handling.In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility, and suitable experiments have been carried out aiming to deeply investigate the pool thermal-hydraulic behavior of a HLM cooled pool reactor.In particular a fuel pin bundle simulator (FPS) has been installed in the CIRCE pool. It has been conceived with a thermal power of about 1 MW and a linear power up to 25kW/m, relevant values for a LMFR. It consist of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The pins have a diameter of 8.2mm and active lengths of 1 m. Along the FPS, three spacer grid properly designed by ENEA have been installed.The FPS has been deeply instrumented by several thermocouples. In particular three sections of the FPS have been instrumented to monitor the heat transfer coefficient along the bundle as well as the cladding temperature in different rank of sub-channels.A full characterization of the FPS has been experimentally achieved both under forced and natural circulation, and the main results gained during the run are reported into the paper.Moreover the paper reports a preliminary analysis and discussion of such results, also in comparison with CFD calculations performed by CFX code.Copyright