DEMO Water-Cooled Lithium Lead Research Articles

RELAP5-based thermal-hydraulic assessment of the STEAM facility for DEMO WCLL balance of plant analysis

DEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328–295 °C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238–300 °C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up.STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions.

STEAM Experimental Facility: A Step Forward for the Development of the EU DEMO BoP Water Coolant Technology

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations.

Numerical study of MHD mixed convection flow in the EU DEMO WCLL breeding blanket

The Water-Cooled Lithium Lead (WCLL) breeding blanket is a proposed candidate option for European DEMO nuclear fusion reactor. The utilization of PbLi as the predominant material occupying the bulk region of the breeding blanket has received significant attention. The interaction between Lorentz forces (Ha≈9800), intense buoyancy forces (Gr≈1.77×1010), and pressure-driven flow (Re≈236.77), plays a crucial role in the magnetohydrodynamics(MHD) and heat transfer phenomena of the region. To replicate the dynamics of PbLi, taking into account the impact of the strong magnetic field used for plasma confinement and the non-uniform heat source generated by nuclear heating, a 3-D DNS numerical simulation has been employed. Owing to the influence of buoyancy forces, several medium-sized circulations that are elongated in the direction of the magnetic field occupy the plenum region which is situated between the cooling pipes. Moreover, the electrical interaction between the liquid metal and the conducting wall gives rise to high-velocity flow jets, which exhibits peak velocities that are increased with the magnitude of the magnetic field. To characterize the heat transfer occurring in different regions of the system, local Nusselt and Grashof numbers have been determined. Optimization strategies for the BZ cooling system geometrical configurations are proposed based on the estimation of heat transfer and prediction of tritium permeation.

Thermomechanical and Thermofluid-Dynamic Coupled Analysis of the Top Cap Region of the Water-Cooled Lithium Lead Breeding Blanket for the EU DEMO Fusion Reactor

In the EU, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept is one of the candidates for the design of the DEMO reactor. From the past campaign of analysis emerged that the thermal-induced stress led to the failure in the verification of the RCC-MRx structural criteria. Hence, in this paper the classic conceptual design approach, based on a pure FEM thermal and structural analysis, is compared to a coupled thermofluid-dynamic/structural one. Even though the coupled approach requires tremendous modelling effort and computational burden, it surely allows determining the thermal field with a higher level of detail than the FEM analysis. Therefore, in this work, the focus is put on the impact of a more detailed thermal field on the DEMO WCLL BB global structural performances, focusing on the Top Cap region of its Central Outboard Blanket segment. The obtained results have allowed confirming the soundness of the design solution of the Top Cap region, except for concerns arising on the mass flow rate distribution. Moreover, results have shown that, globally, the pure FEM approach allows for obtaining more conservative results than the coupled one. This is a positive outcome in sight of the follow-up of the DEMO WCLL BB design, as it will be still possible adopting the pure FEM approach to quickly down-select design alternatives, using the most onerous coupled approach to finalise the most promising.

Water distillation for coolant purification system of DEMO water-cooled lithium lead breeding blanket

During the DEMO pre-concept design phase, a review of existing and under development technologies suitable for the Coolant Purification System (CPS) of the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) was carried out and two tritium recovery strategies were individuated, named on-line and off-line. The on-line concept relies on the continuous by-pass and treatment of a certain primary coolant fraction into CPS whereas the off-line scheme foresees the discharge of the whole primary water inventory after one year of DEMO operation, that will be treated in an external facility inside the tritium plant. Among these, the on-line strategy has been recently recommended as the most suitable for lowering the tritium inventory within BB and Primary Heat Transfer System (PHTS). Furthermore, such solution does not need high decontamination factor since treated water is not discharged in the environment, but it is redirected again into PHTS. In such condition, Water Distillation (WD) could be a promising technology with intrinsic simplicity and safety. The present work deals with the feasibility assessment of WD for DEMO WCLL BB CPS, investigating its capability to meet water CPS duty. A parametric study has been preliminary carried out, analysing the effect of selected parameters (i.e., tritium concentration at CPS inlet, tritium concentration in the enriched stream, efficiency of anti-permeation barriers, and operative conditions) on the WD sizing and operation. Then, a preliminary reference design has been individuated and verified with a dynamic model developed in MATLAB environment. Finally, the process flow diagram of the system has been derived and a preliminary sizing of auxiliary components is provided.

CFD Analysis and Optimization of the DEMO WCLL Central Outboard Segment Bottom-Cap Elementary Cell

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is achieved by the first wall (FW) and breeding zone (BZ) cooling systems, which in the water-cooled lithium–lead (WCLL) BB employ pressurized water. Different arrangements of BZ coolant conduits have been investigated in the recent past to identify an efficient layout, which could meet the structural materials’ operational temperature constraint and which could provide the optimal coolant outlet temperature. However, most of the computational fluid dynamic (CFD) analyses that have been carried out until now have been focused on the equatorial WCLL elementary cell of the central outboard segment (COB). The aim of this work is to broaden the analysis to other relevant locations in the blanket. An assessment of the design of the cooling system of the COB bottom-cap elementary BZ cell has been identified as a top design priority due to its different geometry and thermal loads. The cooling efficiency of the BZ and FW systems is investigated to assess if the coolant-appropriate design conditions are matched and the temperature distribution in the cell is analyzed to identify the onset of hot spots. Different layouts of the FW systems are proposed and compared in terms of thermal–hydraulic reliability.

Thermo-Mechanical Analysis and Design Update of the Top Cap Region of the DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment

Within the framework of the EUROfusion research and development activities, the Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) is one of the two candidates to be chosen as the driver blanket for the European DEMO nuclear fusion reactor. Hence, an intense research work is currently ongoing throughout the EU to develop a robust conceptual design able to fulfil the design requirements selected at the end of the DEMO pre-conceptual design phase. In this work, the thermo-mechanical analysis and the design update of the top cap (TC) region of the DEMO WCLL Central Out-board Blanket (COB) segment is presented. The scope of the work is to find a design solution of the WCLL COB TC region able to fulfil the design requirements, prescribed by the reference RCC-MRx code, under the selected nominal and accidental steady state loading scenarios. The activity herein presented moved from the WCLL COB reference design, purposely modified in compliance with the adopted thermal and mechanical requirements in order to attain a robust TC region geometric layout. In the end, a geometric configuration called “TC region-mod++” was determined, foreseeing a TC able to safely withstand both nominal and accidental loads. Nevertheless, some criticalities still hold in the internal stiffening plates and, therefore, further and finer analysis are necessary to fully match the goal. In any case, it was also found that the proposed approach for the design update is promising and worthy to be further pursued. The work was performed following a theoretical–numerical approach based on the finite element method (FEM) and adopting the quoted Ansys commercial FEM code.

Analysis of the Thermo-Mechanical Behaviour of the EU DEMO Water-Cooled Lithium Lead Central Outboard Blanket Segment under an Optimized Thermal Field

Within the framework of the EUROfusion research activities on the DEMO Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) design, a research study was performed to preliminarily optimize, from the thermal point of view, the WCLL Central Outboard Blanket (COB) segment in order to investigate its structural behaviour under a realistic thermal field. In particular, a study of thermal analyses was performed to optimize the Double Walled Tubes and Segment Box cooling channels’ geometric configurations along the poloidal extension of the WCLL COB segment, in order to obtain a spatial temperature distribution fulfilling the thermal design requirement. Then, the thermo-mechanical analysis of the WCLL COB segment under Normal Operation (NO, representing nominal conditions), Upper Vertical Displacements Event (UVDE, representing a plasma disruption event) and Over-Pressurization (OP, representing an in-box loss of coolant accident) scenarios were carried out, assuming the previously obtained thermal field, to realistically predict displacement and stress fields. Finally, a stress linearization procedure allowed comparing the stress values obtained in some critical regions of the structure with the criteria prescribed by the reference design standard RCC-MRx. A theoretical–numerical approach based on the Finite Element Method (FEM) was followed using the commercial code Abaqus v. 6.14.

Structural Assessment of the European DEMO Water-Cooled Lithium Lead Breeding Blanket Central Outboard Segment Under Remote Maintenance Loading Conditions

Within the framework of the European research activities devoted to design the DEMO breeding blanket (BB), the Water-Cooled Lithium-Lead (WCLL) BB concept is one of the candidates currently assessed in EU. One of the main issues connected to its design is represented by the remote maintenance (RM) operations. In particular, previous thermo-mechanical analysis highlighted a Ductile to Brittle Transition Temperature (DBTT) shift above the room temperature within the WCLL central outboard (COB) segment during normal operation. This may pose problems during the RM operations foreseen in the BB replacement phase. Indeed, within those WCLL COB segment’s regions where the DBTT shifted above the room temperature, the stress level arising during RM operations may be able to promote crack growth and migration resulting in a fragile fracture event. Then, in this work, the structural assessment of the DEMO WCLL COB segment under RM scenarios is performed, in order to provide the fracture mechanics analysis with the necessary input in terms of stress spatial distribution and to understand if the deformation occurring during RM can be accommodated by the available space reservation. Several potential RM scenarios have been considered, taking into account also the potential occurrence of a seismic event or the necessity to remove the WCLL COB segment with the liquid breeder not drained. The study has been performed adopting a numerical approach based on the Finite Element Method and adopting the quoted Abaqus code. The obtained results are quite encouraging and will allow, in the follow up, the refinement of the RM plan foreseen for the WCLL BB.

Multiphysics Optimization for First Wall Design Enhancement in Water-Cooled Breeding Blankets

The commercial feasibility of the first fusion power plant generation adopting D-T plasma is strongly dependent upon the self-sustainability in terms of tritium fuelling. Within such a kind of reactor, the component selected to house the tritium breeding reactions is the breeding blanket, which is further assigned to heat power removal and radiation shielding functions. As a consequence of both its role and position, the breeding blanket is heavily exposed to both surface and volumetric heat loads and, hence, its design requires a typical multiphysics approach, from the neutronics to the thermo-mechanics. During last years, a great deal of effort has been put in the optimization of the breeding blanket design, with the aim of maximizing the tritium breeding and heat removal performances without undermining its structural integrity. In this paper, a derivative-free optimization method named “Complex method” is applied for the design optimization of the European DEMO Water-Cooled Lithium Lead breeding blanket concept. To this purpose, a potential performances-based objective function, focusing on the maximization of the tritium breeding, is defined and a multiphysics numerical model of the blanket is developed in order to solve the coupled thermo-mechanical problem, while the optimization algorithm leads the design towards a minimum optimum point compliant with the prescribed requirements. Once the optimized design is obtained, its nuclear and thermo-structural performances are assessed by means of specific neutron transport and multiphysics simulations, respectively. Finally, the structural integrity is verified by means of the application of the RCC-MRx design criteria. • A tool for First Wall optimization in water-cooled breeding blankets is developed. • The optimization is performed using the gradient-free method named “Complex method”. • An objective function aiming at tritium breeding maximization is defined. • A FE model is used to check the compliance with the thermo-structural requirements. • The optimized design is finally assessed by proper, more detailed numerical analyses.

DEMO WCLL primary heat transfer system loops activated corrosion products assessment

The estimation of Activated Corrosion Products (ACPs) in fusion technology is an important input to safety assessments (e.g. occupational radiation exposure, accidental analyses). The present work aims to calculate ACPs for the First Wall (FW) and Breeder Zone (BZ) water cooling loops of the DEMO Water Cooled Lithium Lead (WCLL) fusion reactor. The code used is PACTITER v2.1. Geometrical and thermal-hydraulic inputs have been built from the design and operating data of the DEMO-WCLL.Since an extensive library of experimental data for EUROFER97 corrosion in water is not available yet, correlations obtained from previous calculations have been employed and compared to measured values.The impact of operational scenario and water chemistry conditions is here shown. Preliminary results confirm the preference towards a water chemistry of pH = 7.0 at 300 °C. The continuous scenario provides the most conservative results in terms of mass and activity inventory for ACPs.

Structural assessment of the EU-DEMO WCLL Central Outboard Blanket segment under normal and off-normal operating conditions

Within the framework of the EUROfusion design activities concerning the EU-DEMO Breeding Blanket (BB) system, a research campaign has been carried out at the University of Palermo with the aim of investigating the structural behaviour of the DEMO Water-Cooled Lithium Lead (WCLL) Central Outboard Blanket (COB) segment. The assessment has been performed considering three different loading scenarios: the Normal Operation (NO), the Over-Pressurization (OP) and the Upward Vertical Displacement Event (VDE-up). In particular, NO scenario represents the loading case referring to the nominal operating conditions, whereas the OP scenario refers to the loading conditions due to an in-box LOCA accident, listed as one of the BB design basis accidental events. Lastly, the VDE-up scenario is an off-normal event reproducing the plasma disruption caused by an uncontrolled vertical motion of the plasma volume. This event brings the plasma in contact with the upper part of the plasma chamber, generating a sudden energy discharge accompanied by Electro Magnetic (EM) forces acting on the structure. The study has been carried out following a theoretical-numerical approach based on the Finite Element Method (FEM) and adopting the quoted ABAQUS v. 6.14 commercial FEM code. In particular, a detailed 3D FEM model of the whole COB segment, including the back-supporting structure and its attachment system to the vacuum vessel, has been set up. Several simulations have been run to assess the thermo-mechanical performances of the segment under the afore-mentioned loading scenarios, also taking into account the impact of the tungsten (W)-armour on the overall structural response. EM loads have been considered in all the assessed scenarios. In the first two, only magnetization forces have been taken into account, while in the VDE-up scenario Lorentz’s forces have been also taken into account. The structural response has been evaluated according to the RCC-MRx structural design rules. The obtained results are herewith presented and critically discussed.

Design and preliminary analyses of the new Water Cooled Lithium Lead TBM for ITER

In the European strategy, DEMO is the intermediate step between ITER and a commercial fusion power plant. In this framework, one of the goal of DEMO is to be a Breeding Blanket test facility. The Breeding Blanket, which is not present in ITER, is one of the key components for the future deployment of nuclear fusion electricity as it accomplishes the functions of tritium breeding and nuclear to thermal power conversion.Due to time constraints lead by the construction schedule of DEMO, a new strategy to consider in DEMO a “driver” Breeding Blanket that needs limited technological extrapolation has been chosen, while “advanced” Breeding Blanket concepts will be tested in the next phases. In this context, ITER will allow to test technologies to provide relevant contributions in terms of Return of eXperience to the DEMO “driver” Breeding Blanket project by the mean of Test Blanket Modules (TBM) to be installed in different ITER Vacuum Vessel Ports.Among the possible “driver” Breeding Blanket, the Water Cooled Lithium Lead (WCLL) concept comes out. In this framework, a realignment of the DEMO Breeding Blanket and TBM programs has started in 2017, leading to a new TBM development relevant of the DEMO WCLL BB. The WCLL TBM is therefore an essential component in ITER that will provide crucial information for the development of the DEMO “driver” blanket.This paper aims at presenting the development process and design status of WCLL TBM. After recalling the main features of the WCLL TBM, conceptual design analyses are presented and discussed.

Thermal-hydraulic analysis of the DEMO WCLL elementary cell: BZ tubes layout optimization

Abctract The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling systems: First Wall (FW) that faces the plasma heat flux and the Breeding Zone (BZ) that removes the deposited power of neutron and photon interaction. Several layouts of WCLL BB system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and suitable thermal-hydraulic performances. This research activity focuses on the equatorial WCLL elementary cell of the Central Outboard Segment (COB), based on the WCLL BB 2018 design of DEMO 2017 baseline, investigating the cooling performances of the BZ and FW systems and their mutual interaction, verifying the reliability to deliver the coolant at adequate design conditions (i.e. 328 °C), evaluating the Eurofer temperature field in order to optimize the BZ Double Wall Tubes (DWTs), extinguishing the hot spot onset into the structures due to the implementation of the IAEA PbLi thermal conductivity as reference. The goal of the study is to compare three specific different tubes arrangement of the BZ system: the V0.6, that is the reference configuration, the V0.6_A and V0.6_B, identifying and discussing advantages and key issues from the thermal-hydraulics point of view. The analyses have been carried out using a CFD approach, adopting the commercial ANSYS CFX code, thus a 3D finite volume model of each configuration has been developed. For each WCLL design, several steady-state analyses have been performed, to evaluate the Eurofer temperature field and the thermal-hydraulics performances under different cooling configuration, changing the position of the DWTs. The results show that the FW and BZ system have a mutual interaction. Only the V0.6_B layout can safely remove the high heat loads deposited into the elementary cell, respecting the DEMO requirements. This research activity aims at laying the basis for the finalization of the WCLL BB design.

Parametric study of the influence of double-walled tubes layout on the DEMO WCLL breeding blanket thermal performances

Abstract Within the framework of the EUROfusion activities regarding the EU-DEMO Breeding Blanket (BB) concept, the University of Palermo is long-time involved, in close cooperation with ENEA, in the design of the Water Cooled Lithium Lead (WCLL) BB, that is one of the two concepts under consideration for the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder and neutron multiplier, as well as by subcooled pressurized water flowing as coolant under PWR-like conditions (pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C). A research campaign has been recently carried out to study the potential influence of the Breeding Zone cooling circuit layout, composed of bundles of Double Walled Tubes (DWTs), on the blanket thermo-mechanical behaviour. In the first phase, a parametric study has been performed to investigate the effects of some geometrical DWT parameters onto the blanket thermal response in order to identify the best tubes layout, in case arranged along toroidal-radial planes. In the second phase, the thermo-mechanical performances of the blanket equatorial region, equipped with the optimized DWTs layout, have been assessed. The study has been carried out following a theoretical-numerical approach based on the Finite Element Method and adopting the ABAQUS v. 6.14 commercial code. Model, assumptions and results obtained are herewith presented and critically discussed.

Preliminary design of the top cap of DEMO Water-Cooled Lithium Lead breeding blanket segments

Within the framework of EUROfusion R&D activity, a research campaign has been carried out at the University of Palermo, in close cooperation with ENEA labs, in order to preliminary design the top cap foreseen for the DEMO Water-Cooled Lithium Lead (WCLL) breeding blanket segments. Due to the high heat and pressure loads acting on such component, its design results particularly demanding and a specific multi-physics approach is needed, covering several aspects from thermal-hydraulics to structural assessments. Preliminary detailed CAD model of the cap integrated into the upper region of the WCLL breeding blanket outboard central segment has been set-up, equipped with proper cooling circuits as well as manifold and attachment systems according to the design of the equatorial elementary cell. A detailed numerical model has been set-up with the aim of simulating the thermo-mechanical behaviour of the above-mentioned system. Finally, the thermo-mechanical response of the upper breeding blanket region has been evaluated in terms of stress and temperature distributions, verifying that the structural material maximum temperature stays below its limit value and that structural integrity is ensured by means of the fulfilment of design rules reported in RCC-MRx structural design code.

Optimization of the first wall cooling system for the DEMO WCLL blanket

• Thermal-hydraulic analyses of WCLL FW cooling systems in different poloidal position. • FW systems optimization to reduce the number of water channels, not exceeding the temperature limit of 550 °C. • Steady-state analyses to evaluate the cooling capability of the optimized systems. • Transient analysis to demonstrate the fully withstanding of the DEMO thermal loads. • Thermal-hydraulic aspects of FW designs are discussed and CHF evaluated. The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling systems: First Wall (FW) that faces the plasma heat flux and the Breeding Zone (BZ) that removes the deposited power of neutron and photon interaction. Several layouts of WCLL BB system have been investigated in the last years to identify a configuration that guarantees Eurofer temperature below the limit (550 °C) and good thermal-hydraulic performances. This research activity focuses on the FW cooling system based on the WCLL BB 2018 design of DEMO 2017 baseline, investigating the cooling performances in order to optimize the FW design, reducing the amount of water that affects the Tritium Breeding Ratio (TBR) and improves the efficiency of the Primary Heat Transfer System (PHTS), verifying the reliability to deliver coolant at adequate design temperature (328 °C) to the PHTS. Different solutions have been considered and analyzed focusing on three specific positions along the poloidal direction: the equatorial cells in the Central Outboard Segment (COB) and Inboard Segment (IB), where there is the maximum deposited power and low Heat Flux (HF), and the apical cell of the IB, impacted by the highest HF but with low deposited power. For each FW design, several thermal-hydraulic steady-state analyses have been performed, as well as sensitivity analyses to evaluate the response of the systems under different cooling configuration, changing the position and the number of channels. Furthermore, a transient analysis reproducing the Dwell-Pulse phase of the DEMO fusion reactor with the optimized configuration of the Outboard FW layout has been performed to study the behavior of the FW cooling system subjected to the pulsed operation. The research activity aims at laying the basis for the finalization of the WCLL BB design, pointing out relevant thermal-hydraulic aspects. The analyses have been carried out using a CFD approach, thus a 3D finite volume model of each configuration has been developed, adopting the commercial ANSYS CFX code. The goal of the study is to compare the different FW cooling system layouts, identifying and discussing advantages and key issues from the thermal-hydraulic point of view. The results show that the coolant system of FW can safely remove the high plasma HF and nuclear heat deposition with a reduction of water channels, delivering coolant to the FW PHTS heat exchanger at the design temperature.

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.

Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept

Abstract In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Development of a WCLL DEMO First Wall design module in the SYCOMORE system code interfaced with the neutronic one

The pre-conceptual design of the DEMOnstration reactors has already started and several tokamak configurations have to be tested to find the best design by exploring different design parameters. Fast simulations involving the different components behavior must be performed. Within the European framework, SYCOMORE (SYstem COde for MOdelling tokamak REactor) is developed by CEA for this purpose. The Breeding Blanket (BB) facing the plasma is a key component in DEMO ensuring tritium self-sufficiency, shielding against neutrons and heat extraction for electricity production. Several BB concepts are being studied, among which the Water Cooled Lithium Lead (WCLL) one. SYCOMORE includes several specific modules in Python linked together, one of which has been developed to define a suitable design of the WCLL Breeding Blanket and is presented in this paper.The method to define automatically the WCLL First Wall (FW) design using analytical design formulae starting from thermo-hydraulic and thermo-mechanical considerations as well as design criteria coming from Codes & Standards (C&S) is presented. Furthermore WCLL FW design obtained with SYCOMORE is compared to Finite Elements (FE) analyses of the DEMO WCLL BB.Finally, a coupling between thermo-mechanics and neutronics is implemented, several iterations are necessary to obtain a converged design. Neutronic block evaluates the radial build, the BB tritium production, and the nuclear heating in the FW and the Breeding Zone (used by thermo-mechanical block). Thermo-mechanical module gives the design data (FW thickness, compositions, etc.) to the neutronic block.