Transient analysis of a locked rotor/shaft seizure accident involving the EU-DEMO WCLL Breeding Blanket primary cooling circuits

The EU-DEMO Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) main subsystems to be cooled are the Breeder Zone (BZ) and the First Wall (FW). Each subsystem will be equipped with an independent Primary Heat Transfer System (PHTS). Within the framework of the EUROfusion Work Package Breeding Blanket research program, several accidents belonging to the category of “Decrease in Coolant System Flow Rate” were studied. The activity was aimed at evaluating the blanket and primary cooling systems thermal-hydraulic performances during such transient conditions. A complete model including the BB and related PHTS circuits has been developed at Sapienza University of Rome. A modified version of RELAP5/Mod3.3 system code has been used to perform the calculations. The simulation results showed that a locked rotor/shaft seizure of a BZ or a FW main coolant pump is the most challenging scenario. BZ and FW system behavior has been analyzed following this initiating event with the goal of the design improvement and to individuate the need for preventive measures. The influence of loss of off-site power on the accident evolution has also been investigated. Moreover, management strategies have been proposed for different reactor components. Calculations demonstrate that the current blanket and PHTS design is appropriate to cope with these kinds of accident scenarios

Preliminary neutron kinetic. Thermal hydraulic coupled analysis of the ALFRED reactor using PHISICS/RELAP5-3D

The development of a lead-cooled fast reactor (LFR) demonstrator was proposed, mainly in EU, to investigate the feasibility of an industrial size ELFR (European Lead-cooled Fast Reactor). The demonstrator, called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator), consists of a pool-type lead-cooled fast reactor, with a nominal thermal power of 300 MWt. This paper aims to verify the capability of the PHISICS/RELAP5-3D coupled approach to simulate transients of such reactor and to evaluate the effects of accidental scenarios relevant for the safety analysis on the system thermal-hydraulics and on the core power spatial evolution. RELAP5-3D©, developed at Idaho National Laboratory (INL), is a thermal-hydraulic system code, validated for a wide range of transient simulations. The code provides the possibility to simulate innovative working fluids (such as lead and lead alloys) and to use a fully integrated multi-dimensional nodalization. In addition, the need to study complex neutronic systems recommended the development of a new kinetic model allowing the calculation with any number of energy groups and also considering the transport for the angular variations. At this purpose, INL developed PHISICS (Parallel and Highly Innovative Simulation for INL Code System) and its coupling methodology with RELAP5-3D. The simulation activity described in this paper has been focused on the safety analysis of ALFRED reactor assuming the occurrence of two unprotected transient scenarios: unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP). At this purpose, a thermal-hydraulic geometrical scheme of the whole reactor has been developed. The models and the outcomes of the calculations are described and discussed in the paper, highlighting the capability of the coupling approach to obtain results consistent with the ones available in the literature

Tokamak cooling systems and power conversion system options

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work

Thermal-hydraulic transient analysis of the FFTF LOFWOS Test #13

In the framework of a Coordinate Research Project (CRP) endorsed by the International Atomic Energy Agency (IAEA), the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome has developed a thermal–hydraulic model of the Fast Flux Test Facility (FFTF). FFTF was a Sodium-cooled Fast Reactor (SFR) developed by the Westinghouse Electric Corporation for the U.S. Department of Energy, which operated from 1980 to 1992. The main mission of the reactor was to prove feasibility and operation of SFR nuclear power plants. For this purpose, FFTF was designed as a flexible system and extensively instrumented, making it attractive for code validation purposes. The proposed benchmark exercise involves a Loss Of Flow Without Scram (LOFWOS) test, performed in 1986. DIAEE has developed and is validating a multiphysics modelling based on a Neutron-Kinetic/Thermal-Hydraulic (NK/TH) coupling approach with RELAP5-3D© and PHISICS codes. The present paper deals with the assessment of the standalone TH modelling, establishing the basis for the next NK/TH phase. Steady-state results have been compared with benchmark specifications, proving the adequacy of the model calibration. After that, the comparison between transient simulation and experimental data has highlighted an overall good agreement, demonstrating the capability of the developed TH nodalization to reproduce the FFTF LOFWOS Test #13. Outside of the parameters selected for the comparison in the benchmark exercise, outlet plenum thermal stratification and sodium free surface motion within Gas Expansion Modules (GEM) have been analysed, providing useful outcomes for the understanding of the transient event and for future perspectives

Study of the EU-DEMO WCLL breeding blanket primary cooling circuits thermal-hydraulic performances during transients belonging to LOFA category

The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes

Transient analysis of OSU-MASLWR with RELAP5

The present paper deals with the assessment of the original and a modified version of RELAP5/MOD3.3 against the OSU Multi Application Small Light Water Reactor (OSU-MASLWR). The new implemented features regard suitable correlations for the heat transfer coefficient evaluation in helical geometry. Furthermore, two different modelling of the Helical Coil Steam Generator (HCSG) are assessed. In the first approach, HCSG's primary and secondary sides are collapsed in a single pipe component. In the second model, three equivalent pipes are conceived for the simulation of the three ranks composing the HCSG. Two different power manoeuvring experiments are reproduced. The simulations highlight a satisfactory agreement in both the transients. Nevertheless, the modified code shows enhanced capabilities in the prediction of the HCSG operation. This is due to the improvements adopted in the modified version of RELAP5/MOD3.3 that allows a better modelling of the dryout phenomena occurring within helical tubes, as well as a better estimation of the primary side heat transfer coefficient. The better agreement of the heat exchange is propagated to the primary system, resulting in a more accurate prediction of the inlet and outlet core temperatures, and primary flow rate

Analysis of EU-DEMO WCLL Power Conversion System in Two Relevant Balance of Plant Configurations: Direct Coupling with Auxiliary Boiler and Indirect Coupling

Among the Key Design Integration Issues (KDIIs) recently selected for the DEMOnstra-tion Fusion Power Plant (DEMO), the operation of the Balance of Plant (BoP) Power Conversion System (PCS) has been recognized as a crucial aspect, due to the typical pulsed regime characteriz-ing the fusion power plant. In the framework of the DEMO Water-Cooled Lead-Lithium Breeding Blanket (WCLL BB) concept, three BoP solutions have been recognized to be able to overcome this issue. They rely on different coupling options between the Primary Heat Transfer Systems (PHTSs) and the PCS: an Indirect Coupling Design (ICD) with Intermediate Heat Transport System (IHTS) and Energy Storage System (ESS), a Direct Coupling Design (DCD) with AUXiliary Boiler (AUXB), and a DCD with small ESS. The present paper deals with a preliminary feasibility assessment of the first two solutions. The analysis, carried out with the GateCycle™ code, referred to a preliminary design phase, devoted to the sizing of the main components, and to a second phase focused on the cycle optimization. The study demonstrated the feasibility of the two BoP concepts. They are able to produce a satisfactory average electric power (>700 MW) with an acceptable average net electric efficiency (33.6% for both concepts). For each solution, the main strengths and weaknesses are com-pared and discussed

Evaluation of the thermal-hydraulic performances of a once-through steam generator in nuclear fusion applications

After decades of operation in nuclear power plants, Once-Through Steam Generators (OTSGs) were recently proposed for nuclear fusion applications. In particular, they are supposed to be installed in the primary cooling systems of the European Union Demonstration fusion power plant (EU-DEMO). One of the key reactor components is the Breeding blanket (BB). Among the BB concepts that are currently under study, Water-Cooled Lithium-Lead (WCLL) option was considered for this work. The WCLL blanket is divided in two main subsystems, the breeder zone (BZ) and the first wall (FW), each one provided with an independent cooling circuit, named Primary Heat Transfer System (PHTS). Thermal power removed from BB by BZ and FW PHTS is driven to the Power Conversion System (PCS) to be converted into electricity. The thermal coupling is ensured by two OTSGs per system. At the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome, a simulation activity was carried out to understand the component thermal-hydraulic behavior during DEMO normal operations. For calculation purposes, a full model of the steam generator was prepared by using a modified version of RELAP5/MOD3.3 system code. The computational activity performed allows to preliminary characterize the OTSG thermal-hydraulic performances during both pulse and dwell phases

Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis

In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients. A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action