MHD mixed convection flow in the WCLL: heat transfer analysis and cooling system

In the Water-Cooled Lithium Lead (WCLL) blanket, a critical problem faced by the design is to ensure that the breeding zone (BZ) is properly cooled by the refrigeration system to keep the structural materials under the maximum allowed temperature by the design criteria. CFD simulations are performed using ANSYS CFX to assess the cooling system performances accounting for the magnetic field effect in the sub-channel closest to the first wall (FW). Here, intense buoyancy forces (Gr = 10^10) interact with the pressure-driven flow (Re = 10^3) in a MHD mixed convection regime. A constant magnetic field, parallel to the toroidal direction, is assumed with intensity B = 4.4 T. The walls bounding the channel and the water pipes are modeled as perfectly conducting. The magnetic field is found to dampen the velocity fluctuations triggered by the buoyancy forces and the flow is similar to a forced convection regime. The PbLi heat transfer coefficient is reduced to one-third of its ordinary hydrodynamic value and, consequently, hot-spots between the nested pipes and at the FW are observed, where TMax = 1000K. Optimization strategies for the BZ cooling system layout are proposed and implemented in the CFD model, thus fullling the design criterion

DEMO WCLL BB breeding zone cooling system design: analysis and discussion

The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium self-sufficiency, shielding the Vacuum Vessel and removing the heat generated in the tokamak plasma. The last function is fulfilled by the First Wall (FW) and Breeding Zone (BZ) independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years in order to identify a configuration that guarantee Eurofer temperature below the limit (823 K) and good thermal-hydraulic performances (i.e. water outlet temperature 601 K). A research activity is conducted to study and compare four configurations, which rely on different arrangement of the stiffening plates (i.e. toroidal-poloidal and radial-poloidal), orientation of the cooling pipes (i.e. horizontal, vertical) and PbLi flow path. The analysis is carried out using a CFD codes, thus a threedimensional finite volume model of each configuration is developed, adopting the commercial ANSYS CFX code. The objective is to compare the BZ cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view, also considering feedbacks from MHD and neutronics analyses. The research activity aims at laying the groundwork for the finalization of the WCLL blanket design, pointing out relevant thermal-hydraulic aspects

Implementation of the chemical PbLi/water reaction in the SIMMER code

The availability of a qualified system code for the deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. Considering the renewed interest for the WCLL breeding blanket, such code shall be multi-phase, shall manage the thermodynamic interaction among the fluids, and shall include the exothermic chemical reaction between lithium-lead and water, generating oxides and hydrogen. The paper presents the implementation of the chemical correlations in SIMMER-III code, the verification of the code model in simple geometries and the first validation activity based on BLAST Test N°5 experimental data

Qualification of TRACE V5.0 Code against Fast Cooldown Transient in the PKL-III Integral Test Facility

The present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry caused by an increase of the heat transfer from the primary to the secondary side that induces a fast cooldown transient on the primary side-affected loop. The asymmetric overcooling effect requires an assessment of the reactor pressure vessel integrity considering PTS (pressurized thermal shock) and an assessment of potential recriticality following entrainment of colder water into the core area. The aim of this work is the qualification of the heat transfer capabilities of the TRACE code from primary to secondary side in the intact and affected steam generators (SGs) during the rapid depressurization and the boiloff in the affected SG against experimental data

BEST ESTIMATE CODES UNCERTAINTY EVALUATION IN THE NUCLEAR TECHNOLOGY: IMPLEMENTATION AND DEVELOPMENT OF CIAU METHODOLOGY TO CATHARE 2 CODE (VALUTAZIONE DELL’INCERTEZZA NEI RISULTATI DEI CODICI BEST ESTIMATE APPLICATI ALLA TECNOLOGIA NUCLEARE: IMPLEMENTAZIONE E SVILUPPO DELLA METODOLOGIA CIAU PER IL CODICE CATHARE 2)

Il presente programma di dottorato si pone come obiettivo generale l’approfondimento delle problematiche concernenti la valutazione dell’incertezza nei codici per analisi di "Best Estimate", contribuendo allo sviluppo e incrementando le capacità della procedura CIAU, e rendendo la stessa applicabile alle predizioni effettuate con il codice del CATHARE2 sviluppato in Francia presso il CEA(Commissariat à le Energie Atomique). Nell’introduzione, dopo aver descritto i vari tipi di approccio per effettuare le analisi di sicurezza, tre sezioni sono dedicate alla rilevanza che le analisi “Best Estimate” hanno per l’ottimizzazione del “design” dei nuovi impianti e del funzionamento dei vecchi a fronte dei vincoli imposti nel processo di “licensing”. Il Capitolo 1 dedicato agli obbiettivi della presente tesi di dottorato e ad una dettagliata descrizione di come tali obbiettivi sono stati raggiunti e nell’ambito di quali progetti. Il secondo capitolo è dedicato ad una descrizione dei vari metodi di incertezza applicabili ai codici termoidraulici “BestEstimate”. La procedura CIAU è estensivamente descritta nel Capitolo 3, fornendo adeguate nozioni delle definizioni e dei principi su cui si basa. Il Capitolo 4 è diviso in due parti: una prima parte è dedicata ad un’introduzione del codice di riferimento dell’attività, il CATHARE2 (F); la seconda alla realizzazione del database di accuratezze e incertezze ottenute attraverso analisi effettuate su apparecchiature sperimentali. L’ultimo capitolo (il Capitolo 5), prima delle conclusioni (Capitolo 6), è dedicato alla qualifica e all’applicazione della procedura CIAU attraverso l’utilizzazione del database sviluppato e descritto nel Capitolo 4

Prediction of Void Fraction in PWR Subchannel by CATHARE2 Code

The current generation of thermal-hydraulic system codes benefits of about sixty years of experiments and forty years of development and are considered mature tools to provide best estimate description of phenomena and detailed reactor system representations. However, there are continuous needs for checking the code capabilities in representing nuclear system, in drawing attention to their weak points, in identifying models which need to be refined for best-estimate calculations. Prediction of void fraction and Departure from Nucleate Boiling (DNB) in system thermal-hydraulics is currently based on empirical approaches. The database carried out by Nuclear Power Engineering Corporation (NUPEC), Japan addresses these issues. It is suitable for supporting the development of new computational tools based on more mechanistic approaches (i.e. 3 field codes, 2 phase CFD, etc.) as well as for validating current generation of thermal-hydraulic system codes. Selected experiments belonging to this database are also used for the OECD/NRC PSBT benchmark. The paper presents the validation activity performed by CATHARE2 v2.5_1 (six equation, two field) code on the basis of the sub-channel experiments available in the database and performed in different test sections. Four sub-channel test sections are addressed in different thermal-hydraulic conditions (i.e. pressure, coolant temperature, mass flow and power). Sensitivity analyses are carried out addressing nodalization effect and the influence of the initial and boundary conditions of the tests

Hybrid 1D + 2D Modelling for the Assessment of the Heat Transfer in the EU DEMO Water-Cooled Lithium-Lead Manifolds

The European demonstration fusion power reactor (EU DEMO) tokamak will be the first European fusion device to produce electricity and to include a breeding blanket (BB). In the framework of the design of the EU DEMO BB, the analysis of the heat transfer between the inlet and outlet manifold of the coolant is needed, to assess the actual cooling capability of the water entering the cooling channels, as well as the actual coolant outlet temperature from the machine. The complex, fully three-dimensional conjugate heat transfer problem is reduced here with a novel approach to a simpler one, decoupling the longitudinal and transverse scales for the heat transport by developing correlations for a conductive heat-transfer problem. While in the longitudinal direction a standard 1D model for the heat transport by fluid advection is adopted, a set of 2D finite elements analyses are run in the transverse direction, in order to lump the 2D heat conduction effects in suitable correlations. Such correlations are implemented in a 1D finite volume model with the 1D GEneral Tokamak THErmal-hydraulic Model (GETTHEM) code (Politecnico di Torino, Torino, Italy); the proposed approach thus reduces the 3D problem to a 1D one, allowing a parametric evaluation of the heat transfer in the entire blanket with a reduced computational cost. The deviation from nominal inlet and outlet temperature values, for the case of the Water-Cooled Lithium-Lead BB concept, is found to be always below 1.4 K and, in some cases, even to be beneficial. Consequently, the heat transfer among the manifolds at different temperatures can be safely (and conservatively) neglected

Capabilities of TRANSURANUS Code in Simulating BWR Super-Ramp Project

After one-two years of normal operation in a LWR, the fuel–cladding gap may close, as a result of as a result of several phenomena and processes, including the different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction). In this equilibrium state, a significant increase of local power (like a transient power ramp, i.e. power increase in the order of 100kW/m-h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold, material dependent, cracks typical of stress corrosion may appear and grow-up: this phenomenon is called stress corrosion cracking (SCC). The cracks of the cladding may spread out from the internal surface, causing the fuel failure. The objective of the activity (performed in the framework of the IAEA CRP FUMEX III), is to validate the TRANSURANUS models relevant in predicting the fuel failures due to PCI/SCC during power ramps. Focus is given on the main phenomena, which are involved or may influence the cladding failure behavior. The database selected is the Studsvik BWR Super-Ramp Project, which belongs to the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments (IFPE) database” by OECD/NEA. It comprises the data of sixteen BWR fuel rods, that have been modeled and simulated with suitable input decks. The burn-up values range between 28 and 37 MWd/kgU. Eight rods, of KWU standard type, are subjected to fast ramps, the remaining rods experience slow ramps and are of standard GE type

GEN-IV LFR development: Status & perspectives

Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR. In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA. CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied. NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection. LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed. All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development

Parametric thermal-hydraulic analysis of the EU DEMO Water-Cooled Lithium-Lead First Wall using the GETTHEM code

The system-level code GETTHEM is applied to the thermal-hydraulic analysis of an entire segment of the Water-Cooled Lithium-Lead (WCLL) First Wall (FW) of the EU DEMO reactor, parametrically varying the heat load of the FW and the coolant mass flow rate. The results show that the WCLL FW design can tolerate variations of the distribution of the heat flux with respect to the design value, without requiring modifications. The top inboard and the bottom outboard regions are identified as most critical from the point of view of the cooling of the FW. Finally, the largest possible extent of the WCLL FW surface where the peak heat load can be safely applied is identified through a parametric analysis, performed on the critical regions, to understand which is the limit of the cooling capacity of the system