1,721,026 research outputs found
Design and thermal-hydraulic transient analysis of primary cooling systems for tokamak fusion reactors
Full text linkThe 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
Validation of RELAP5-3D© for liquid metal reactor technologies
Full text linkThe present research work set in an international and national context that includes the efforts of several universities and research centers in a strict collaboration. Within the international framework, the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of “Sapienza” University of Rome (UNIROMA1) has been recently involved in the Horizon 2020 (H2020) SESAME (thermal-hydraulic Simulations and Experiments for the Safety Assessment of MEtal-cooled reactors) project. The project aims to contribute to the liquid metal-cooled fast reactors (LMFRs) development, including the advanced numerical approaches for the design and safety evaluation of the technologies. Regarding the national context, R&D efforts are mainly dedicated on the development of the lead-cooled fast reactor (LFR) technologies, involving three main partners: ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), CIRTEN (Interuniversity Consortium for Technological Nuclear Research) and the industrial partner ANSALDO NUCLEARE. Strong collaboration and comparing among the three partners are devoted to the development of the reference LFR project: the Advanced Lead Fast Reactor European Demonstrator (ALFRED).
In this framework, the purpose of the present research activity has been to contribute to the understanding of relevant thermal-hydraulic phenomena that characterize the operations of the LMFR, and to the fundamental validation process of the innovative numerical tools, adopted for safety analysis and licensing of new Generation IV (GEN IV) technologies. The research activity has dealt with the validation of RELAP5-3D© (R5-3D) for applications on liquid metal-cooled pool-type fast reactors.
The thesis is divided in seven chapters. The first one is dedicated to the definition of the general background of the research activity. Chapters 2, 3 and 4 are focused on the validation of R5-3D for application on LMFRs. The merits of the simulations are evaluated comparing the results with experimental data from CIRCE (CIRColazione Eutettico) facility, Phénix reactor and PERSEO (in-Pool Energy Removal System for Emergency Operation) facility. The numerical activities have been used to explore and validate different modelling approaches and the acquired know-how has been used to support the design of ALFRED reactor. This topic is examined in chapter 5 that is basically divided in two subsections: the first one analyzes the reference configuration of ALFRED and the second one deals with the revised configuration of the reactor. Chapter 6 presents a methodology for the uncertainty quantification, based on the coupling approach between RELAP5-3D and RAVEN codes. Experimental data from the loop-type facility, called NACIE (NAtural CIrculation Experiment), have been used for the qualification of the methodology. Finally, the main results and guidelines, as well as the weaknesses and the future perspective coming out during the present research activity, are pointed out in chapter 7.
CIRCE is a multipurpose pool-type facility aimed to investigate thermal-hydraulics of innovative heavy liquid metal (HLM) cooled pool-type systems. Two experimental campaigns have been considered in the present work, related to two different configurations of the facility: ICE (Integral Circulation Experiment) and HERO (Heavy liquid mEtal pRessurized water cOoled tubes).
The experimental campaign promoted in CIRCE-ICE test facility was aimed to investigate the thermal-hydraulics of a complex HLM system and to provide data for validation of computational tools. Two experimental tests have been analyzed in this activity: Test A, consisting in a transition from no-power to full power steady-state conditions, and Test I, consisting in a transition from gas-enhanced circulation (GEC) to natural circulation (NC), simulating a protected loss of heat sink plus loss of flow accident. The computational activity has been addressed to investigate the capability of RELAP5-3D© to predict thermal stratification phenomenon in an HLM pool. Several examples were found in literature concerning the simulation of large tanks with RELAP5. The state of art on the simulation of the thermal stratification in large pool has been confirmed by the calculations performed on CIRCE-ICE. The mono-dimensional approach, using a single channel for pool modelling, highlights discrepancies with the experimental data, failing on the prediction of the axial temperature profile. A high peak temperature in the middle of the tank, which is not observed by the experiment, is caused by a total absence of the natural flow within the pool. In order to verify how the axial conduction within the fluid can improve the computational results, a thermal conduction model has been implemented in the nodalization, using several heat structures that couple adjacent meshes. As expected, the axial conduction reduces the peak but not enough to match the experimental temperature profile. To reproduce buoyancy within the LM tank, the pool has been divided in three vertical channels, connected with cross junctions. This approach (model #1) has shown good capabilities on the thermal stratification evaluation. The qualitative trend is well reproduced, predicting two relevant stratifications in the upper and in the middle volumes of the pool. In the lower part of the tank, the LBE temperature is very well simulated but, at 4 m, the plateau temperature is under-predicted of about 15 K and this discrepancy is maintained up to the cover gas. R5-3D was improved with a fully integrated multi-dimensional (MULTID) modelling scheme, mainly developed for volumes where the movement of the fluid is preferably 1D. The MULTID component has been used for the simulation of CIRCE pool. In addition, the fuel pin simulator (FPS) modelling has been improved for the subchannel analysis (model #2). Several figures of merit have been selected, assessing the capabilities of the two models to reproduce thermal-hydraulics of an HLM cooled pool-type system in safety-relevant operations. The comparison with experimental data has highlighted excellent capabilities of the two models to predict thermal-hydraulics of the main flow path, managing to evaluate the most important features: LBE mass flow rate in both GEC and NC conditions, heat exchange within FPS, heat exchanger (HX) and decay heat removal system (DHR), and heat losses. In addition, model #2 assesses RELAP5-3D abilities as a subchannel analysis code, in both GEC and NC operations. The effect of the radial conduction has been evaluated: specific heat structures have been implemented to reproduce thermal conduction through adjacent subchannels. This analysis has shown small effects of the radial conduction, even if, in low flow rate regimes, such as NC operation, it provides not negligible improvement on the prediction of the temperature profile. In this case, the simulation is in good agreement with experimental data; the highest discrepancies are observed in the edge of the bundle (4 degrees) where the errors can be justified by the uncertainties related to the thermocouples positions. Focusing on the pool simulation, the MULTID component has introduced relevant improvements on the prediction of the thermal stratification phenomenon. The two relevant stratifications have been observed by the calculation. The temperature in the lower part matches very well the experimental measurements. The lower stratification level is well predicted and the temperature hot plateau underestimation has been reduced to 5 K. After the transition from GEC to NC, both the models are able to predict the upper stratification attenuation and the movement of the lower one below the DHR outlet. In the long term, the two calculations provide the same temperature profile that matches well the experimental trend, limiting the discrepancies below 4 degrees. Evaluation of the axial conduction within the pool has been performed, highlighting limited effects, especially when the natural circulation (NC) inside the pool is considered.
HERO test section was employed in CIRCE facility to investigate thermal-hydraulics of a double wall bayonet tube (DWBT) steam generator (SG), in a relevant configuration for ALFRED SG. Moreover, a validation benchmark was proposed within the H2020 SESAME project to asses the capabilities of different computational tools to predict the main thermal-hydraulic phenomena in an HLM-cooled pool-type facility. UNIROMA1 supported the definition and the realization of the experimental campaign, developing a nodalization scheme, based on the validated CIRCE-ICE model, for the pre-test analysis. The calculations have investigated different transient scenarios, highlighting the capabilities of HERO test section to guarantee sufficient NC to remove the decay heat in the short term. Based on these results, a set of three experimental tests has been performed, consisting in three protected loss of flow accidents (PLOFAs), occurred during the normal operation of the facility. The numerical activity has concerned two of the three tests, adopting an improved nodalization scheme. Regarding post-test analysis of Test 3, simulation of the full power conditions is globally in agreement with experimental data for all the primary circuit physical quantities monitored, including the thermal stratification phenomenon; some discrepancies are highlighted on the secondary side, mainly due to the lack of some information which determines large uncertainties on the boundary conditions related to the operation of the secondary loop. Starting from the full power steady-state conditions, two transient calculations have been performed, assuming the same boundary conditions, except for the feedwater (FW) mass flow rate after the transition event. Case 1 assumes the reference value of the secondary flow rate (0.078 kg/s), obtained with the energy balance equation applied to the FW pre-heater. This calculation highlights an overestimation of the power removed by the SG. A second calculation has been performed adjusting the total secondary flow rate to 0.047 kg/s; that value guarantees the correct SG power. Both the simulations are in good agreement with experimental data in the first 200 s, reproducing very well the first minutes after the transition event. After the Ar injection cut off, the first calculation provides a good estimation of the minimum value of the LBE MFR, underpredicted by the second calculation of about 2 kg/s. The long term behavior strongly depends on the feedwater mass flow rate. Case 1 shows an overestimation of the whole system energy unbalance, leading to the overprediction of the natural circulation contribution and of the cooling trend. The SG power balance analysis has highlighted an overestimation of the power removed of about 30% of the experimental value. The large uncertainties related to the measurement of the secondary system quantities have suggested the calibration of the secondary flow rate to obtain the correct SG power removed. This assumption has been justified comparing the experimental flow rate, acquired at the inlet of the tubes 0 and 4, with the simulation results: the experimental data are underestimated but the calculation results remain within the experimental uncertainty bands. Looking at the primary system, the assumption of a lower FW flow rate leads to a better agreement with experimental data, providing a good estimation of the long term behavior. Some discrepancies are still maintained on the secondary side, where the steam outlet temperature is overpredicted by the code. The differences could be due to a not perfect agreement of the powder thermal conductivity, which represents a large source of uncertainties. This opens the possibility to continue the post-test analysis for the secondary side, leading an improved experimental results analysis, despite the good global results obtained. Similar conclusions are obtained from the post-test analysis of Test 1, confirming good capabilities of R5-3D to reproduce thermal-hydraulics of HLM-cooled systems in safety-relevant operations. In addition, the subchannel analysis performed has highlighted a good prediction of the subchannel thermal-hydraulics within the FPS bundle in the postulated transient accident.
The Phénix Dissymmetric End-of-Life test, proposed for a benchmark exercise within the H2020 SESAME project, offered useful data for the analysis of more complex systems where asymmetrical effects could play a relevant role in the system thermal-hydraulics. As a participant to the validation benchmark, in collaboration with ENEA, UNIROMA1 has developed a detailed nodalization of the reactor, including a three-dimensional modelling of the pools and an assembly per assembly core modelling in the active region, suitable for a neutron kinetic and thermal-hydraulic (NK-TH) coupling calculation. The full power calculation has highlighted a good capability of the code to reproduce the normal operation of the reactor. Starting from the steady-state results, the transient calculation has been performed assuming the dissymmetric test boundary conditions provided by CEA. The asymmetric distribution of the flow rate through the secondary loops leads to an asymmetric operation within the primary system, which is well predicted by R5-3D. The asymmetrical operation of the two secondary systems leads to a dissymmetric evolution of the thermal-hydraulics within the cold pool. Good agreements have been observed between experiment and simulation. In particular, the movement of the hot sodium within the cold pool is well predicted by the code, which is able to predict the local peak temperature at the PP1 inlet. At this regard, the three-dimensional momentum equation adopted in the MULTID modelling seems to provide a good instrument for the evaluation of the temperature and flow distribution within large volumes.
Safety and reliability are relevant aspects of the development of GEN IV reactors. In this framework, LMFRs present peculiar characteristics related to the thermophysical properties of the coolant, basically regarding the possibility of coolant freezing, that can occur in long term DHR operation if the thermal power removed by the DHR system exceeds the decay residual power. For this reason, in LMFRs, the DHR system must ensure an efficient power removal, avoiding to overcome technological limits in terms of maximum temperatures, and must prevent coolant freezing in the grace time period. In addition, according to GIF guidelines, passive DHR systems are needed to prevent unexpected evolution of accidental scenarios following a total loss of the continuous electrical power supply. The solution consisting of an isolation condenser (IC) immersed in a water tank, acting as a final heat sink, could meet the above-mentioned characteristics. The operation of such a system, is based on in-tube condensation under NC condition and pool boiling. For this reason, a validation process is required for STH codes, such as R5-3D. In this framework, PERSEO facility provided useful experimental data. UNIROMA1 developed a 1D model of the facility. In order to reduce the computational cost, a mono-dimensional model of the pools (three parallel pipes with cross junctions) has been included. This modelling choice is justified by the negligible effect of the thermal stratification on the system thermal-hydraulics. The numerical activity has shown satisfactory capabilities of the code to reproduce the safety performance of the passive system. The main limitation observed in R5-3D calculations, and in almost all the STH codes adopted in the benchmark exercise, is the significant underestimation of the power exchanged between the HX and the HX pool. This can be attributed to the underprediction of the heat transfer coefficient (HTC) in both the tube-side and pool-side, where the condensation under natural circulation conditions and the pool boiling are outside of the validity ranges of the correlation fully integrated into the code. For this reason, the main improvement adopted in the nodalization has been the application of a constant multiplicative factor (2.4) to the HTC on both the sides.
The main target of the validation activity was to qualify R5-3D STH code to support the design of the GEN IV nuclear power plants. The numerical activity was used to explore and validate different modelling approaches and the acquired know-how has been used to support the design of ALFRED reactor, investigating the reference and the improved configurations of the LFR concept. The first numerical activity has concerned the reference concept of the reactor, developed within the LEADER project. Based on the experience learned during the analysis of the experimental campaign performed in CIRCE and on the references found in literature, a thermal stratification phenomenon was expected into the main pool of the reactor. For this reason, a detailed three-dimensional model of the pool has been developed. In addition, an assembly per assembly core modelling, assuming the approach adopted for Phénix simulations, has been developed, allowing the calculation of the power distribution at the beginning of life (BOL) of the reactor using the NK-TH methodology. It is based on the RELAP5-3D/PHISICS coupling calculation. The full power calculation has highlighted a relevant thermal stratification, of about 70 K, in the upper part of the pool, that is not been involved in the primary flow path. Thermal stratification represents a significant technological issue. This was one of the reasons that encourages the designers to develop a revised concept of ALFRED reactor. In this frame, the solution was to include an internal structure, within the Reactor Vessel, that forces the cold lead, exiting the SG, to move upward and then, passing through specific holes in the upper part of the IS, to move downward towards the core inlet. In this way, zones not involved in the flow path are avoided. A thermal-hydraulic model of the new configuration was developed in order to verify, among other issues, the absence of relevant thermal stratifications in both normal and accidental operations. For this purpose, a detailed MULTID component was developed to reproduce the pools of the reactor. The numerical activity has demonstrated the improved pool thermal-hydraulics: significant thermal stratifications are avoided in both full-power operation and SBO scenario. Another issue is related to the possibility of coolant freezing in all the operative conditions, including accidental scenarios. A solution that limits the maximum temperatures in the first instants after the transition event and modulates the power removed from the primary coolant in the long term, has been proposed, adopting an IC, immersed in water pool, equipped with non-condensable (nitrogen) tank. The non-condensable tank allows to passively modulate the power removed by the DHR system, by injecting nitrogen within the IC bundle. The non-condensable flow rate is passively controlled by the pressure difference between the gas tank and the water system. The numerical activity has been aimed to verify the revised concept of the DHR system under a postulated protected loss of offsite power (PLOOP). The simulation shows the capability of the safety system to restrict the maximum temperatures in the first phases of the transient within the technological limits. In the long term, as the depressurization of the secondary system occurs, nitrogen is injected within the IC bundle, degrading the heat exchange. This is enough to limit the power removed by the DHR system to the decay heat value, limiting the Pb minimum temperature to 630 K, about 30 K higher the Pb freezing point.
Finally, the last chapter of the thesis has proposed a best estimate plus uncertainty (BEPU) methodology, based on a statistical exploration of the input space considering the associated uncertainties altogether and the analyses of the responses with several validation metrics. The methodology consists in the RELAP5-3D/RAVEN coupled calculation. The objective has been to verify the coupled methodology for the uncertainty quantification of LM-cooled systems. For this purpose, the experimental campaign performed on NACIE facility and the thermal-hydraulic model developed by UNIROMA1 has been considered. The UQ has been based on the perturbation of the input space following a Monte Carlo sampling, propagating the input uncertainties. The analysis of the main outcomes related to selected FOMs, has been performed with three comparison metrics, fully integrated into RAVEN tool. The application of the comparison metrics has shown the capabilities of the methodology, highlighting the merits and the weaknesses of the thermal-hydraulic model. The future perspectives could be the application of the model to more complex models to increase the validation process of the methodology and to apply it to the verification process of the new NPP concepts.
The main outcomes and guidelines coming out during the research activity are summarized in four main sections:
1. Pool modelling:
• Several modelling approaches have been explored;
• Mono-dimensional approach based on a single equivalent channel fails to reproduce large pool thermal-hydraulics;
• Mono-dimensional modelling approach, consisting in more parallel channels connected with cross junction, is preferable when satisfactory prediction of pool thermal-hydraulics are required, even if phenomena such as thermal stratification do not assume relevant role in the behavior of the whole system. This modelling approach allow to reduce computational cost;
• MUILTID modelling approach is preferable if relevant t
Verification of TRANSURANUS Code Versions v1m1j07 and v1m1j08 against BWR Inter-Ramp Experiments
No full textTechnical Repor
Post Test Analisys by Relap5 Code & Accuracy Quantification of PKLIII E3.1 Experiments (Sensitivity calculation)
No full text
Thermal hydraulic design of DEMO Water Cooled Lithium Lead Breeding Blanket and integration with primary system and balance of plant
Full text linkThe Ph.D. work, conducted at ENEA C.R. Brasimone, is carried out in the framework of the European Power Plant and Physics Technological Programme, under the coordination of the EUROfusion Consortium, and it was co-funded by the EUROfusion Engineering Grant.
The aim of this Ph.D. thesis work is the development of the conceptual design of the Water Cooled Lithium Lead breeding blanket and its Primary Heat Transfer System, as well as their integration, demonstrating the compatibility with the DEMO requirements.
The activity is focused on the thermal-hydraulic design, the sizing and analyses of the Breeding Blanket system and of the main components of the Primary Heat Transfer System, Energy Storage System and Power Conversion System. This has been pursued through engineering approaches and the application of numerical tools, such as thermal-hydraulic system codes and CFD codes.
The conceptual design of WCLL blanket was developed starting from a review of previous designs. A preliminary layout of the coolant systems (first wall and breeding zone) and the main parameters were defined, through engineering tools, to provide input data for the development of the CAD model.
In order to verify the thermal-hydraulic performances of the WCLL blanket system, a complete three-dimensional finite volume model of the breeding blanket was set-up, using ANSYS CFXv15.0 code. The model includes solid and fluid domains, and represents, in detail, an elementary cell of the blanket (i.e. breeding unit). CFD analyses have been carried out to investigate thermal and fluid-dynamic behavior of the breeding blanket, evaluating the efficiency of the first wall and breeding zone water coolant systems, and identifying key issues and areas of improvements. Several configurations were analyzed, and it was identified a promising coolant system layout which ensures a symmetric thermal field of the breeding blanket with maximum temperature of solid structures of 430 °C.
One of the main functions of the breeding blanket is the conversion of the energy coming from the plasma in thermal energy suitable for power generation systems, ensuring an efficient power conversion. It requires large external auxiliary systems to perform its function, namely the Primary Heat Transfer System and the Power Conversion System, and it has to be integrated with them in a complete Balance of Plant, which satisfies DEMO tokomak constraints and requirements. This objective is pursued, for the WCLL breeding blanket, in the second part of the research activity, investigating the postulated operation of DEMO power plant through thermal hydraulic analyses of the Primary Heat Transfer System and Power Conversion System design solutions and components. Moreover, considering the pulsed nature of DEMO reactor, with energy generated for 120 min (burn time) followed by the reactor dwell time (estimated to last 10-30 min), an Intermediate Heat Transfer System equipped with an Energy Storage System, is being investigated to mitigate the impact of plasma pulsing on PCS equipment (e.g. in the steam turbine) and in electrical grid.
The research activity led to the definition of the Primary Heat Transfer System and of the DEMO WCLL BB. The selected configuration relies on two separate systems connected with the breeding zone and the first wall, respectively. The Energy Storage System is foreseen, to accumulate energy during pulse time, using HITEC molten salt as fluid, and to deliver power to the Power Conversion System during dwell time. The main components (e.g. steam generators, circulators/ pumps, pipes, collectors) were sized, and the data were used to develop and integrate the CAD model into the DEMO baseline. A preliminary Gate-cycleTM analyses were carried out, presenting an average gross electrical efficiency of about 37.1%, considering both pulse and dwell phases.
In order to develop a dynamic model of the systems, an extended version of RELAP5/Mod3.3 code was set-up with the implementation of the PbLi and HITEC fluid properties. This had allowed to develop a thermal-hydraulic system model of the first wall and breeding zone primary systems. The model includes the in-vessel and ex vessel components of the primary side and the secondary side of the FW PHTS and BZ PHTS, and it will be used to perform thermal-hydraulic system analyses
Integrated Analysis for a Small Break LOCA in the IRIS Reactor Using MELCOR and RELAP5 Codes
No full text
Going Beyond Counting First Authors in Author Co-citation Analysis
Full text linkThe present study examines one of the fundamental aspects of author co-citation analysis (ACA) - the way co-citation
counts are defined. Co-citation counting provides the data on which all subsequent statistical analyses and mappings
are based, and we compare ACA results based on two different types of co-citation counting - the traditional type that
only counts the first one among a cited work's authors on the one hand and a non-traditional type that takes into
account the first 5 authors of a cited work on the other hand. Results indicate that the picture produced through this non-traditional author co-citation counting contains more coherent author groups and is therefore considerably clearer. However, this picture represents fewer specialties in the research field being studied than that produced through the traditional first-author co-citation counting when the same number of top-ranked authors is selected and analyzed. Reasons for these effects are discussed
- …
