Conceptual Design of the Steam Generators for the EU DEMO WCLL Reactor

In the framework of the EUROfusion Horizon Europe Programme, ENEA and its linked third parties are in charge of the conceptual design of the steam generators belonging to EU DEMO WCLL Breeding Blanket Primary Heat Transfer Systems (BB PHTSs). In particular, in 2021, design activities and supporting numerical simulations were carried out in order to achieve a feasible and robust preliminary concept design of the Once Through Steam Generators (OTSGs), selected as reference technology for the DEMO Balance of Plant at the end of the Horizon 2020 Programme. The design of these components is very challenging. In fact, the steam generators have to deliver the thermal power removed from the two principal blanket subsystems, i.e., the First Wall (FW) and the Breeding Zone (BZ), to the Power Conversion System (PCS) for its conversion into electricity, operating under plasma pulsed regime and staying in dwell period at a very low power level (decay power). Consequently, the OTSG stability and control represent a key point for these systems' operability and the success of a DEMO BoP configuration with direct coupling between the BB PHTS and the PCS. In this paper, the authors reported and critically discussed the FW and BZ steam generators' thermal-hydraulic and mechanical design, the developed 3D CAD models, as well as the main results of the stability analyses and the control strategy to be adopted

Design and thermal-hydraulic transient analysis of primary cooling systems for tokamak fusion reactors

The PhD activity discussed in this document was conducted between 2018 and 2021. It profited from a collaboration between the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome and the Experimental Engineering Division of ENEA at Brasimone. Within the framework of EUROfusion Consortium research activity, the R&D efforts focused on the investigation of one principal blanket option for the European DEMOnstration Power Plant (EU-DEMO): the Water-Cooled Lead-Lithium (WCLL). For this concept, ENEA and its Italian related partners are the principal investigators. During last years, DIAEE played an important role in the conceptualization of the WCLL Breeding Blanket (BB) and its related primary cooling systems. In addition, an extended transient analysis was carried out to assess their thermal-hydraulic performances in both normal operations and accidental conditions. Such work was carried out involving research activities related to both International Thermonuclear Experimental Reactor (ITER) and EU-DEMO fusion power plant. This document is articulated in seven sections. The first one defines the PhD activity framework. In order to perform system-level transient analysis of tokamak reactors, a modified version of the thermal-hydraulic code RELAP5/Mod3.3 was developed at DIAEE in collaboration with ENEA. The aim is enhancing the code modelling capabilities with respect to fusion power plants. Section 2 is dedicated to discuss the implemented features. Sections 3 and 4 refer to the research activity involving DEMO WCLL. In § 3 the pre-conceptual design of the blanket component and related primary cooling circuits is described in detail. Their thermal-hydraulic model, developed for calculation purposes, is treated in § 4. The same section also reports the outcomes of the transient analysis. In the same way, § 5 and 6 are related to ITER WCLL-Test Blanket System (TBS) research activity. The TBS conceptual design, in particular the one of Water Cooling System (WCS) circuit whose DIAEE is responsible for, is described in § 5. To perform the system thermal-hydraulic assessment a dedicated model was developed. Its detailed description is provided in § 6, together with a full comment of the calculation results. Finally, § 7 focuses on the main conclusions and future perspectives of the work done. The first issue to be addressed was the development of a suitable code to perform the computational activity. System thermal-hydraulic codes are the reference numerical tools adopted for the nuclear reactor transient analysis. Most of them, such as RELAP5, were developed and validated to perform best-estimate transient simulations of Light Water Reactors (LWR). Once validated against experimental data coming from more than one-hundred facilities, they have been used throughout decades to perform the licensing of LWRs. Simulation results allowed to characterize the reactor transient behavior in the full range of operative and accidental conditions. The same approach to reactor transient analysis was envisaged also for fusion power plants. Although, existing system codes lack of some specific features required to properly simulate the fusion reactor performances. For this, during the last years, a modified version of the system code RELAP5/Mod3.3 was developed at DIAEE, including some new upgrades needed to address the modelling issues arising from the simulation of tokamak fusion reactors. New implementations consist in: i) lead-lithium (PbLi) and HITEC© working fluids, with their thermophysical properties; ii) new heat transfer correlations for liquid metals and molten salts; iii) helicoidally tubes dedicated heat transfer correlations and two-phase flow maps. The effectiveness of the new features introduced was verified throughout the three years of research activity by performing transient simulations involving tokamak reactors. Referring to the helicoidally geometry, the new two-phase flow maps were also tested against experimental data coming from OSU-MASLWR (Oregon State University - Multi Application Small Light Water Reactor) facility. In particular, a power manoeuvring test (named ICSP Test SP3) was selected for benchmarking purposes. Several power steps of the Fuel Pin Simulator, standing for the reactor core, was reproduced, from 80 to 320 kW. The aim of the experiment was to investigate the primary system natural circulation and secondary system superheating for a variety of core power levels and feedwater flow rates. The effects of the code modifications on the simulation outcomes were clearly visible at higher power levels when the heat transfer within the HCSG plays a more important role. Indeed, above a certain power threshold, nearly 200 kW, the default version showed limited capabilities to reproduce the test. On the contrary, the trends related to the modified version fit quite well the experimental data. Regarding the DEMO WCLL, in this document, it was presented the outcome of the pre-conceptual design developed during the just finished Horizon 2020 research programme. The design activity performed at DIAEE which the candidate took part to was mainly related to the BB Primary Heat Transfer System (PHTS) layout. The main system function is to remove the heat produced in the blanket components, delivering such thermal power to the Power Conversion System (PCS) to be converted into electricity. The BB PHTS is divided in two independent cooling systems, foreseen for the heat removal from the Breeder Zone (BZ) and the First Wall (FW). Both the BZ and the FW PHTSs consist of two cooling loops based on proven technologies extrapolated from Pressurized Water Reactors (PWR). Each primary system comprises the in-vacuum vessel cooling circuit, the ex-vacuum vessel equipment (pumps, heat exchangers/steam generators and a pressurizer), and the correspondent connecting lines. The BB PHTS is conceived in order to avoid a loop segregation. The BZ/FW PHTS cold legs feed the cold ring, which accomplishes the distribution of the cold water to each in- vacuum vessel cooling circuit (one per each sector). Primary coolant removes power from the blanket components and is collected in the hot ring that delivers water to the hot legs. In case of pump trip in a single PHTS loop, the other cooling loop guarantees the power removal from the whole system after the plasma shutdown. With the aim of the design improvement, system-level transient analyses were run involving the WCLL blanket component and related PHTS. The DIAEE version of RELAP5/Mod3.3 was used for this purpose. Such activity was related to EUROfusion Consortium Work Packages Breeding Blanket (WPBB) and Balance of Plant (WPBoP). Firstly, a full DEMO WCLL thermal-hydraulic model was prepared, considering the BoP Indirect Coupling Design option. Blanket was simulated using equivalent components characterized by lumped parameters. The BZ and FW PHTS circuits were modelled including all the components within and outside the vacuum vessel. PCS nodalization starts from the main feedwater line and arrives up to the Turbine Stop Valves. Thus, only the BZ Once Through Steam Generators (OTSG) secondary side was simulated. Regarding the Intermediate Heat Transfer System (IHTS), the same approach was adopted. Only the cold and hot legs upwards/downwards the FW Heat EXchangers (HEX) shell side were added to the input deck. PCS feedwater and IHTS molten salt conditions at the BZ OTSGs and FW HEXs secondary side inlet were provided by means of boundary conditions. The model developed was tested against the design data by simulating the full plasma power state. Beginning of Life conditions were considered. Proportional-Integral (PI) controllers were implemented to: i) regulate the primary pump rotational velocity and set the required value of the system flow; ii) control the PCS feedwater and IHTS molten salt mass flows in order to obtain the required PHTS water temperature at blanket inlet (i.e. OTSG outlet, 295 °C). Simulation results were in good agreement with the nominal values, demonstrating the appropriateness of the nodalization scheme prepared and of the control system implemented. Blanket and PHTS thermal-hydraulic performances in this flat-top power state were fully characterized, including the calculation of the system pressure drops and heat losses. Then, this steady-state calculation was used as initial condition to perform the DEMO WCLL transient analysis, including some operational and accidental transients. The DEMO reactor normal operations were simulated, including both pulse and dwell phases. Reference plasma ramp-down and ramp-up curves were adopted for simulations purposes. Primary pumps were kept running at nominal velocity for the whole transient, as for DEMO requirement. In addition, during dwell, PHTS circuits must be operated at the system average temperature (nearly 310 °C). Since no control strategies related to BZ OTSGs and FW HEXs were available, a preliminary management strategy for the PCS feedwater and IHTS molten salt mass flows were proposed and investigated. The BB PHTS parameters calculated by the code were analyzed to assess the circuit performances. The imposed trends proved to be effective in keeping the PHTS average temperature during dwell at the required value. After, it was performed a benchmark exercise involving DEMO reactor power fluctuations. System code results were compared with the more detailed ones obtained with ANSYS CFX. The aim was to evaluate the effectiveness of the thermal-hydraulic model developed for the blanket component, prepared using equivalent components characterized by lumped parameters. BZ and FW PHTS water temperatures at blanket outlet were selected as figures of merit. Their trends showed a good agreement between the simulation outcomes obtained with the system code and the Finite Element Method (FEM). Results obtained from this benchmark exercise also indicated an effective way to perform simulations involving components, such as the breeding blanket, characterized by complex geometries and heat transfer phenomena. System code and 3D calculations can be externally coupled in an iterative process where CFX provides more accurate parameters to refine the RELAP5 model and the latter is used to update the inlet conditions for finite volume model computation. Finally, the blanket primary cooling system response during accidental conditions was investigated. The selected transients to be studied belong to the category of “Decrease in reactor coolant system flow rate”. This transient analysis was aimed at understanding the thermal-hydraulic response of the blanket component and related primary circuits. In this way, it was possible to evaluate the appropriateness of their pre-conceptual design and the eventual need of mitigation actions to withstand such accidental scenarios. Different faults that could result in a decrease of the BB PHTS primary flow were postulated and investigated. In particular: i) partial and complete loss of forced primary coolant flow; ii) primary pump shaft seizure (or locked rotor); iii) inadvertent operation of a loop isolation valve. Firstly, the most limiting of the above primary flow decrease event was chosen. It consisted in the complete loss of forced circulation in both FW and BZ PHTS. In this ‘worst case’ scenario, even if very unlikely, a sensitivity was performed on the flywheel to be added to the PHTS main coolant pumps in order to keep the system temperatures within acceptable ranges. The proper moment of inertia values to be applied to BZ and FW primary pumps were selected according to the simulation outcomes. Later, they were also used in all the following transient calculations. The initiating events mentioned above were all simulated when interesting either BZ or FW system components (i.e. pumps and loop isolation valves). Calculations were replicated also considering the influence of loss of off-site power, assumed to occur in combination with the PIE. An actuation logic, involving some components of the DEMO reactor, was proposed and preliminary investigated. It was inspired by the one used for Generation III + nuclear power plants. Results highlighted how the type of circulation (natural or forced) characterizing each cooling system is the main element influencing the correspondent thermal-hydraulic performances. If forced circulation is available, the following behavior can be observed in BZ and FW systems. Few seconds after the start of transient, the temperature spikes at blanket outlet characterizing the trend of both BZ and FW PHTS water are significantly smoothed. In FW system, the availability of forced circulation in both primary and secondary (only for the first 10 s) circuits limits the pressure increase and avoids the intervention of the pressurizer Pilot-Operated Relief Valve (PORV) in the short term. The OTSGs cooling capability lasts less. The presence of forced circulation in the primary cooling system enhances the steam generator heat transfer coefficient, increasing the thermal power transferred to the PCS. This reduces the time between two subsequent steam line Safety Relief Valves (SRV) openings and speeds up the evacuation of the water mass present in the OTSGs secondary side. Once terminated, the steam generators are no more able to provide any cooling function to the BZ PHTS. For more or less two hours from PIE occurrence, the system pressure is controlled by the pressurizer sprays. The first PORV intervention in the long term is significantly delayed. The temperature slope characterizing both BZ and FW systems (thermally coupled) is higher since pumping power is added to the power balance. This is valid until the pump trip is triggered in each system. Summarizing, forced circulation improves the BZ and FW performances in the short term, smoothing the temperature spikes, but reduces the ones in the mid-long term. In fact, it shortens the cooling interval provided to the BZ PHTS by the steam generators and increases the temperature slope experienced by BZ and FW systems, reducing the reactor grace time. The best management strategy for PHTS pumps is to use, at the start of transient, the forced circulation they provide, in order to avoid excessive temperatures in the blanket, and then stop them, to increase the reactor grace time. In all the transient simulations, BZ and FW systems experienced a positive temperature drift in the mid-long term. It is due to the unbalance between decay heat produced in the blanket and system heat losses, with the former overwhelming the latter. The temperature slope is higher if the forced circulation is still active. In these cases, it must be added another source term to the power balance, represented by the pumping power. In the calculations performed, no Decay Heat Removal (DHR) system was implemented in the input deck and the power surplus is managed by the pressurizer PORV. Power in excess produces a pressure increase and when this parameter reaches the PORV opening setpoint, PHTS water mass is discharged with its associated enthalpy content. This is the way adopted by BZ and FW system to dissipate the power surplus. In the future developments of this research activity, the impact of the DHR system will be also evaluated. In conclusion, simulation outcomes highlighted the appropriateness of the current PHTS design and of the management strategy chosen for the selected accidental scenarios. During the third ITER council (2008), it was established the so-called ITER Test Blanket Module (TBM) program. Its objective is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. More recently, in 2018, the WCLL option was inserted among the selected blanket concepts to be investigated. From this time, an intense research activity was conducted within the EUROfusion Work Package Plant level system engineering, design integration and physics integration (WPPMI) in order to perform the pre-conceptual and conceptual design phases of ITER WCLL Test Blanket System. The overall work (i.e. TBS) was divided in ‘Part A’, related to TBM set and ‘Part B’, referring to its related ancillary systems. For the latter, R&D effort was led by ENEA and involved many European research institutions and universities, including DIAEE of Sapienza University of Rome. The work was supervised also by Fusion for Energy, the EU organization managing Europe’s contribution to ITER reactor. By the fall of 2020, both design phases were concluded, and the system successfully underwent its Conceptual Design Review. Among the TBM ancillary systems, the most relevant is the Water Cooling System, acting as primary cooling circuit of the TBM module. The design and thermal-hydraulic characterization of this circuit was up to DIAEE. The TBS conceptual design was presented in this document. A special focus was given to the WCS layout whose DIAEE is responsible for (i.e. the candidate took part to). The Water Cooling System was designed to implement the following main functions: i) provide suitable operating parameters to the water flow cooling the TBM in any operational state; ii) transfer thermal power from WCLL-TBM to CCWS; iii) provide confinement for water and radioactive products; iv) ensure the implementation of the WCLL-TBS safety functions. In addition, ITER WCLL-TBM must be DEMO relevant. Such relevancy refers to the water thermodynamic conditions at the TBM (15.5 MPa, 295-328 °C) since the experimental program will deal with the test of this blanket reference concept. The reduced thermal power produced in the TBM set (near 700 kW) with respect to DEMO blanket (1923 MW), allows to use a single water-cooling system for both the FW and the BZ. The correspondent WCS primary flow was computed considering the TBM power balance. The ultimate heat sink for the WCLL-TBM WCS is the ITER Component Cooling Water System (CCWS). With the aim to include an additional barrier between the contaminated primary water and the CCWS coolant, the WCLL-WCS was split in a primary and a secondary loop. In such a way, the CCWS radioactive inventory is kept below the limit in any operative and accidental scenario (note that CCWS is a non-nuclear system). To simplify the WCLL-WCS management, liquid only condition was selected for the secondary coolant instead of the two-phase fluid, as in DEMO PCS. It is worth to emphasize that electricity generation is not a purpose of ITER and, thus, steam production is not required. CCWS provides water coolant at low pressure and temperature (0.8 MPa at 31 °C), and requires that return temperature must be limited to 41 °C. Hence, there is a considerable difference between the average TBM temperature and the average CCWS temperature. To avoid an excessive temperature excursion (i.e. thermal stresses) between the two sides of a single heat exchanger, an economizer was installed in the middle of the WCS primary loop, leading to the typical “eight” shape of this circuit. Therefore, a total of three heat exchangers were considered for the whole WCS, namely: HX-0001 (economizer), HX-0002 (intermediate heat exchanger between primary and secondary loops) and HX-0003 (heat sink delivering power to CCWS). Each heat exchanger was provided with a bypass line allowing the regulation of the exchanged power by tuning the shell side mass flow. Finally, an electrical heater was added to the WCS primary loop in order to compensate the power unbalance in the system. Most of the WCS equipment is installed in the TCWS Vault, at level four of the tokamak building. The rest of the components, including the TBM, is placed in the level one Port Cell #16. Both locations are linked by means of connection pipes hosted in a vertical shaft. To support the WCS design a preliminary transient analysis was performed. For this purpose, a full thermal-hydraulic model of the system was developed by using the DIAEE version of RELAP5/Mod3.3. Since this circuit is directly connected to PbLi loop within the TBM, also these two systems were included in the overall TBS model. A preliminary control system was implemented for both WCS and PbLi loop. All the main circuit parameters (pressure, temperatures, and mass flows) are controlled in order to ensure system stability in any operative scenario and to provide water coolant and breeder at TBM with the required inlet conditions. Firstly, full plasma power state was simulated at both Beginning of Life (BOL) and End of Life (EOL) conditions. Such calculations were needed to test and evaluate the appropriateness of the model prepared. Simulation outcomes demonstrated that control systems corresponding to WCS and PbLi loop are able to ensure the required values at TBM inlet in both the operative scenarios. For WCS, the main differences between BOL and EOL conditions were highlighted, mainly regarding the operation of the temperature control system (i.e., the mass flow through the heat exchangers bypass). WCS and PbLi loop performances in this flat-top states were fully characterized, including the calculation of pressure and temperature fields, as well as the system power balance. In addition, an insight into the TBM behavior during full plasma power condition was given. Its operation does not change from BOL to EOL since it is provided with water coolant and liquid metal at constant thermodynamic conditions and flow rate. It is important to note that a full thermal-hydraulic characterization of the component was out of the scope of the research activity carried out by DIAEE. Nevertheless, TBM box contains part of the WCS circuit and constitutes the system source term. Furthermore, thermal coupling between W

Subchannel Analysis of LFR Wire-Wrapped Fuel Bundle with RELAP5-3D

A computational campaign was carried out at the Department of Astronautical, Electrical and Energy Engineering of Sapienza University of Rome aiming at the assessment of RELAP5-3D & COPY; capabilities for subchannel analysis. More specifically, the investigation involved a lead-bismuth-eutectic-cooled wire-spaced fuel pin bundle and compared simulation outcomes with experimental data coming from the NAtural CIrculation Experiment-Upgraded (NACIE-UP) facility, hosted at ENEA Brasimone Research Center. Thermal-hydraulic nodalization of the facility was developed with detailed subchannel modeling of the fuel pin simulator (FPS). Three different methodologies for the subchannel simulation were investigated, increasing step by step the complexity of the thermal-hydraulic model. In the simplest approach, the subchannels were modeled one by one. In addition, mass transfer between them was considered thanks to multiple cross junction components, realizing the hydraulic connection between adjacent subchannels. In this case, mass transfer depends on the pressure gradient and hydraulic resistance only, ignoring the turbulent mixing promoted by the wire-wrapped subassembly. Simulation results were not satisfactory, and an improvement was introduced in the second approach. In this case, several control variables calculate at each time step the energy transfer between adjacent control volumes associated with the turbulent mixing induced by the wires. This energy is transferred using ad hoc heat structures (HSs), where the boundary conditions are calculated by the control variables. The present model highlighted good capabilities in the prediction of the radial temperature distribution within the FPS, considerably reducing disagreement with experimental data. Finally, the influence of radial conduction within the fluid domain was assessed, introducing further HSs. Although this most complex model provided the best estimation of the experimental acquisition, the improvements given by radial conduction were not so relevant to justify the correspondent increase of the computational cost

Thermal-hydraulic modeling and analyses of the water-cooled EU DEMO using RELAP5 system code

The conceptual design of the Primary Heat Transfer System (PHTS) of the water-cooled European (EU) DEMO foresees two independent cooling circuits, the breeding zone PHTS and the first wall PHTS. During the pulse time (120 minutes) the first delivers thermal power to the turbine, the latter delivers thermal power to the Intermediate Heat Transfer System (IHTS) equipped by an Energy Storage System (ESS). The IHTS delivers partially thermal power to the turbine in pulse so that to accumulate a suitable amount of energy in ESS to operate the turbine during the dwell time (10 minutes) at almost constant load, despite the EU DEMO pulsation of the generated thermal power. A dynamic model of the primary systems of water-cooled EU DEMO is developed using RELAP5/Mod3.3 system code to verify components sizing and to investigate code predictive capabilities. The model includes the primary and secondary side of the breeding zone and first wall, and in particular: in-vessel (i.e. breeding blanket) and ex-vessel components (i.e. main collectors, hot and cold legs, heat exchangers, steam generators, pumps). Preliminary assessments of the nodalization have been carried out, in particular checking pressure drops along the systems and heat exchanger performances

Development of a Steam Generator Mock-Up for EU DEMO Fusion Reactor: Conceptual Design and Code Assessment

Recent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes

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

The Design of Water Loop Facility for Supporting the WCLL Breeding Blanket Technology and Safety

The WCLL Breeding Blanket of DEMO and the Test Blanket Module (TBM) of ITER require accurate R&D activities, i.e., concept validation at a relevant scale and safety demonstrations. In view of this, the strategic objective of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone of ENEA, is twofold: to conduct R&D activities for the WCLL BB to validate design performances and to increase the technical maturity level for selection and validation phases, as well as to support the ITER WCLL Test Blanket System program. Basically, the Water Loop facility will have the capability to investigate the design features and performances of scaled-down or portions of breeding blanket components, as well as full-scale TBM mock-ups. It is a large-/medium-scale water coolant plant that will provide water coolant at high pressure and temperature. It is composed by single-phase primary (designed at 18.5 MPa and 350 °C) and secondary (designed at 2.5 MPa and 220 °C) systems thermally connected with a two-phase tertiary loop acting as an ultimate heat sink (designed at 6 bar and 80 °C). The primary loop has two main sources of power: an electrical heater up to about 1 MWe, installed in the cold side, downstream of the pump and upstream of the test section, and an electron beam gun acting as a heat flux generator. The WL has unique features and is designed as a multi-purpose facility capable of being coupled with the LIFUS5/Mod4 facility to study PbLi/water reaction at a large scale. This paper presents the status of the Water Loop facility, highlighting objectives, design features, and the analyses performed

Design and Integration of the EU-DEMO Water-Cooled Lead Lithium Breeding Blanket

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed