Dottorato in ENERGIA E AMBIENTE

Titolo della tesi: 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 discussing 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, in collaboration with ENEA, 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 OTSG secondary side. Once terminated, the steam generators are no more able to provide any cooling function.  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. 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. 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 WCS and PbLi loop occurs inside the module. For this, it was mandatory to properly simulate the heat transfer phenomena taking place within the component. The results obtained with the system code were compared with the more detailed ones available in literature (produced by using FEM methodologies). The latter were used to calibrate the component thermal-hydraulic model. Then, the two steady-state calculations were used as initial condition to simulate operative scenarios and abnormal conditions. The Normal Operation State (NOS) was the first to be analyzed. The WCS and PbLi loop control systems were tested to demonstrate their effectiveness in ensuring stable operations against the pulsed regime characterizing the NOS. Both BOL and EOL conditions were considered in order to assess the change in WCS thermal-hydraulic performances with the system aging. The reference ITER pulsed plasma regime was adopted for simulation purposes. The system code results demonstrated the appropriateness of the WCS and PbLi loop control systems. They are able to ensure water coolant and PbLi at the TBM with nearly constant inlet thermodynamic conditions and flow rate. For water inlet temperature, oscillations were limited to +/- 3 °C, acceptable for WCS and TBM operation. Moreover, it was verified that in any part of the PbLi loop an adequate margin (16 °C) from the freezing point is maintained. Finally, in order to assess and verify the WCS design, two abnormal scenarios were selected and investigated. The aim was to evaluate the system capabilities under degraded conditions and to verify if the standard control strategies without any external action are capable to maintain the TBM cooling function for an entire plasma pulse. This last condition allows to avoid the triggering of the Fast Plasma Shutdown System, demonstrating that a minor accident in the WCS does not interfere with the ITER global operation. The transients considered were: i) LOFA occurring in WCS secondary loop; ii) LOHS, i.e. loss of flow in the CCWS. The worst operative condition was supposed to be the EOL, since plugging and fouling limit the heat exchange. For this, NOS at EOL was imposed as initial condition for the transient calculations. In both scenarios, simulation outcomes highlighted that WCS primary loop is kept in safety conditions over the whole accidental evolution. In addition, the safety margin from the PbLi freezing is ensured by keeping the reference water temperature at the TBM inlet. The current WCS design and the control systems implemented proved to be effective to withstand the selected accidental scenarios.