Thermal-hydraulic Analysis Research Articles

Thermal hydraulic transient analysis of ITER chilled water system: Decay heat exchanger leak

Chilled Water System (CHWS-1) is an ITER safety cooling system comprised of two independent and segregated trains (CHWS-1A and CHWS-1B), which provide cooling water to components located in Tokamak and Hot Cell Complexes.VV-PHTS is the only component that, during water baking mode, operates at high pressure than CHWS-1. In case of leaks, it represents a risk in terms of over pressurization and contamination of the CHWS-1 (due to tritiated water and activated corrosion products (ACPs)). Therefore, this paper postulates CHWS-1 leak events at one interface: Vacuum Vessel Primary Heat Transfer System. (VV-PHTS).Three scenarios have been analysed: an incident event (heat exchanger pipe leakage) and two accident events (heat exchanger tube break and multiple tubes break).A simplified RELAP5–3D model of CHWS-1B was developed to perform thermal hydraulic transient analysis as well as to define and size system operation and control.A postulated accident management strategy is included while also defining CHWS-1 instrumentation and control (I&C), which play a relevant role in event detection and mitigation procedures.The simulation results are studied to analyze and verify the system behavior as well as the required actions to reach and maintain a safe state after the given events.With respect to the events studied, this paper reports and confirms the system design to properly and safely bring CHWS-1B to a shutdown, as per corresponding ITER project requirements.This operating scenario has been studied for academic purposes; it is not related to the actual CHWS-1B design. During VV-PHTS baking mode, decay HX is not foreseen to be operative. In order to avoid affecting the redundancy of the system, this study supports this decision.

Conservative analytical and experimental investigation of all pumps failure event in the primary cooling system for a 5 MWth research reactor

The 5-MWth Jordanian Nuclear Research Reactor is an open-tank and open-pool research reactor which uses narrow rectangular fuel. The reactor is cooled by demineralized water in downward forced flow and natural circulation in transient events. The reactor performed a loss of flow test in the Primary Cooling System (PCS) by turning off all PCS pumps at full-power operation. The reactor then displayed safe behavior in thermal-hydraulic, neutronics, and acceptable radiological values. Besides, the RELAP5 MOD3.3 code was used to implement the reactor in three different comprehensive models according to several safety analysis conservatisms, such as assuming a lower total core flow rate by −12% or assuming failure in a safety valve. After that, a benchmark was conducted between the experimental results, and the code calculations of implemented models showed well agreement in general thermal-hydraulic and safety analysis parameters. The code proved a safe reactor trend that used all conservative assumptions even in the most conservative implemented model. The primary safety parameters, the Maximum Temperature of Fuel Clad (MTFC) and the Minimum Critical Heat Flux Ratio (MCHFR), were 421 K and 1.82 for a short time, respectively. The resulting parameters underwent the event's acceptance criteria, where the limits of the set values are 673 K and 1.5 for MTFC and MCHFR, respectively. As a result of the study, several models illustrated differences in thermodynamic parameter effects and gave higher and additional reactor safety levels. Moreover, the test comprehensively investigated the reactor behaviors and ensured the reactor cooling system capability by free convection cooling during the event.

A neutronics optimization approach for preliminary design and safety of nuclear reactors for nuclear thermal propulsion

Nuclear thermal propulsion is a key technology for long-range spaceflight, as demonstrated by the rising interest from space agencies, especially NASA. A preliminary design involves the exploration of many reactor configurations, until a configuration that meets all the design requirements is found. Trade-offs among different design fields, such as neutronics, thermal-hydraulics, safety and rocket performances are unavoidable. Design engineers have to run a large amount of simulations because of the multitude of possible reactor configurations and the different analyses that must be carried out. The present work investigates an optimization approach for the design of a LEU reactor with CERMET fuel, to find a configuration that fulfils neutronics, safety and NASA requirements for a nuclear thermal rocket. A neutronics analysis constitutes the basis of the work, but accidental scenarios and thermal-hydraulics analysis are performed as well, to assess reactor safety and rocket performances. The optimization procedure simulates different configurations and gradually reduces the design space to be explored, until an optimal region is found. In this way, unnecessary simulations are avoided. A Python script is developed to handle the whole analysis, from pre-processing to post-processing, including the integration between the neutronic code Serpent and MATLAB®. Once the optimal region is found, the most promising configurations are identified by comparing different performance metrics retrieved from the neutronics analysis. A safety analysis that simulates the reactor behaviour in four accidental scenarios is carried out on this smaller group of reactor configurations. Finally, reactors that fulfil safety requirements undergo a thermal-hydraulic analysis, which verifies that thermal limits are not exceeded and evaluates rocket performances. The approach successfully reduces the number of configurations to explore, and limits additional analyses to those configurations that are truly competitive. Furthermore, configurations in the optimal region achieve better performance than the initial configuration. It is also found that the most demanding constraints relate to safety: many promising configurations were discarded because they cannot meet safety criteria. This highlights the necessity of including safety analyses in the initial phase of reactor designs. It is concluded that a LEU, CERMET-fuelled reactor design can be considered challenging but feasible, and the approach presented in the work may be applied to find an optimal configuration for a preliminary reactor design.

Comparative study of constitutive relations implemented in RELAP5 and TRACE – Part II: Wall boiling heat transfer

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

Design of a new Small Modular Nuclear Reactor using TVS-2M Fuel Assemblies and Fuel Depletion analysis during the fresh-core cycle length

In this paper, a new fresh Small Modular Nuclear Reactor core with changes on the Fuel Assembly-type, fuel enrichment, and gadolinium (Gd) concentration of the Fuel Assemblies (FAs) is designed and proposed to improve and enhance the NuScale nuclear reactor as a typical PWR-Small Modular Reactor. An algorithm is additionally presented to redesign the core consisting of square lattice FAs into a core including hexagonal FAs. This algorithm is implemented based on the utilization of MCNP code which classifies the design criteria considering simplified modeling and complicated modeling with high computational cost. The algorithm is helpful for a designer to manage the computational cost of modeling during the design process.Core design Basis Limits including Neutronic and Thermal-hydraulics (TH) analysis, are performed and verified for the first cycle of the presented new Small Modular Nuclear Reactor as a Hexagonal NuScale (HNu). Via ANSYS-FLUENT& MCNP codes, TH and Neutronic analysis have been coupled. Also, the burnup calculations for the presented core have been done to analyze the core first cycle length. The excess reactivity and power peaking factor have been assessed during the first-cycle length. Average core fuel depletion has been taken into account, and masses of produced fission fragments and build-up of plutonium isotopes and the remaining mass of uranium isotopes (and Gd) have been computed at the end of the cycle (EOC).The presented new core (HNu) has the benefits of both the TVS-2M FA and NuScale. Compared to the NuScale, HNu has several advantages and superiority, such as increasing MDNBR, decreasing the maximum temperature of fuel and clad, more fuel burn-up, longer working cycle length, and decreasing the maximum power peaking factor.

Pressure drop experiment with externally and internally cooled annular fuel bundles

Canadian Nuclear Laboratories (CNL) has developed an externally and internally-cooled annular fuel (ICAF) bundle that is considered to have improved thermalhydraulic characteristics as compared to the fuel bundles used in the conventional pressurized heavy-water-moderated-and-cooled reactors (PHWRs), such as the CANDU® reactors. The ICAF bundle consists of 24 annular fuel elements (or rods) and a solid central rod. The coolant flows through both the internal and external channels of each annular fuel element; the total flow in the flow tube is split into the flow inside the internal channels of the annular elements and that in the outside subchannels of the fuel bundle. The most compelling argument for this design, compared to the conventional solid-rod design, is the significant reduction in maximum fuel temperature for equivalent linear element ratings, due to significantly-enhanced heat transfer from fuel surfaces to coolant. The primary objective of the ICAF bundle development is to achieve higher burnups and higher power densities using advanced fuel cycles, with comparable or improved thermalhydraulic safety margins. This paper presents the results of a pressure drop experiment performed to investigate the hydraulic behavior of the ICAF bundle. Experimental data were obtained for the pressure distribution, the bundle misalignment junction signature and the flow split. The data can be used to develop predictive models to be implemented into the thermalhydraulics analysis toolset, and to validate the toolset for safety analyses in establishing reactor operating power and evaluating safety margin.

Effect of interfacial area concentration on one-dimensional code simulation of adiabatic two-phase flows in vertical large size channels

In one-dimensional two-fluid model-based codes, an interfacial drag force appears in a momentum conservation equation. The interfacial drag force is an essential parameter in predicting a void fraction. The accurate modeling of the interfacial drag force is indispensable for evaluating thermal–hydraulic characteristics in a nuclear reactor core. The interfacial drag force is formulated as the product of an ‘overall’ drag coefficient and the square of the relative velocity between gas and liquid phases. The ‘overall’ drag coefficient is expressed by the product of a drag coefficient, interfacial area concentration, and density of continuous phase. The rigorous model of the drag coefficient depending on a bubble shape regime has been established. The modeling of the interfacial area concentration depending on a flow regime is one of the weakest links in thermal–hydraulic analysis. The interfacial area transport equation was proposed to predict the dynamic change of the interfacial area concentration but has not reached a level accurate enough to predict the interfacial area concentration. Due to its incomplete development situation, an alternative way to predict the interfacial area concentration through a semi-theoretical interfacial area correlation has been proposed. This study aims to elucidate the effect of the interfacial area concentration on void fraction prediction in a one-dimensional thermal–hydraulic analysis. A one-dimensional two-fluid model-based code, such as TRAC-BF1, has been modified by implementing two existing constitutive equations of the interfacial area concentration into the code. This study also introduces large and small interfacial area concentration models of the interfacial area concentrations. The large and small interfacial area concentration models are designed to intentionally provide hypothetical large and small interfacial area concentrations within physically possible ranges. A total of four interfacial area concentration models is tested under adiabatic two-phase flow conditions. Code calculations with the four different models under steady-state and transient-state conditions and flow regime transitions have identified that the effect of the interfacial area concentration on the void fraction is insignificant for the adiabatic two-phase flows. The findings obtained in this study suggest that a simple interfacial area correlation is sufficient in modeling the interfacial drag force for adiabatic two-phase flows. However, robust and accurate modeling of the interfacial area concentration is still indispensable for two-phase flows with phase change because the interfacial heat transfer term in an energy conservation equation includes the interfacial area concentration.

Pre-conceptual design of an encapsulated breeder commercial blanket for the STEP fusion reactor

As part of the UKAEA Spherical Tokamak for Energy Production (STEP) fusion power station programme, a novel breeding blanket design was assessed. A conceptual design of STEP, which will be an innovative plan for a commercially-viable fusion power station, will be completed by 2024. The final aim of the current assessment is to find a pre-conceptual design that can be manufactured with existing techniques without compromising the fuel self-sufficiency, the heat removal and the shielding functions. The proposed concept, an starting point towards a commercial blanket design, is based on encapsulating the breeding material into hollow spheres forming a gas cooled packed bed. The generated tritium in the breeding material, as well as helium, will will be extracted from the pebbles and directly removed from the blanket by the cooling gas. A frontal multiplier as well as a back reflector have also been studied to maximize the tritium generation. This paper, which is a decided first step towards a breeding blanket for STEP, covers the material design properties and their compatibility, the design configuration, the neutronics analysis, the thermal-hydraulics analysis, a preliminary structural analysis, safety aspects and waste segregation and disposal.

Model development and transient analysis of thermal stratification phenomenon in pool-type sodium-cooled fast reactors

Thermal stratification of pool-type SFR under transient accident conditions has a very significant impact on the performance of key components of the primary loop and the establishment of natural circulation of the residual heat removal system. In this paper, combined with the mechanism of thermal stratification, a three-dimensional (3-D) system analysis model is developed, which takes into account the energy source model, the 3-D convection model and the large space inlet-effect model. Based on the model development, the accuracy of 3-D thermal–hydraulic analysis model and the capability of thermal stratification analysis for SFR were fully verified. The validation results show that the method established in this paper takes into account accuracy and calculation efficiency. It can greatly reduce the amount of numerical grid, take up very little computing resources and greatly improves analyzing efficiency. And it can accurately simulate the thermal stratification phenomenon in the sodium pool. At the same time, the location with maximum temperature gradient of the inner barrel wall can be suggested in the process of thermal stratification. Furthermore, this study provides an important analytical method for the transient safety analysis of SFR, and has a great academic value and engineering significance.

Wall Effect Analysis in Thermal-Hydraulics Aspect of HTR-10 and PR-3000 Reactors Using Different Porosity Models

Gen-IV is the latest generation of a nuclear reactor with better aspects than the previous generation, Pebble-Bed Reactor (PBR) is one of them. PBR using spherical shape fuel or known as pebble fuel. Hence, the physical things that are happening in the core are significantly different from the reactor using a fuel pin. Often the reactor core is assumed as a porous medium because of the void within the pebble fuel. The coolant that moves in the core can be carried as a fluid that flows through a porous medium. The thermal-hydraulics aspect is one of the significant aspects that must have been considered in every nuclear fission reactor because it is closely related to reactor safety. The wall effect happened in a packed bed, making the porosity near-wall position higher than other positions. The porosity value can induce the coolant flow, known as wall channeling, which can influence the reactor’s thermal-hydraulics aspect. In this study, thermal-hydraulics analysis for PBR is performed, and the wall effect is taken into account in the calculations. A thermal-hydraulic code, namely mPEBBLE, was developed based on the PEBBLE code to include the wall effect in the calculation. The code is based on the finite-difference method with axisymmetric cylinder geometry (R-Z) that consists of 4 primary equations; stream flow, pressure recovery, solid energy balance, and thermal energy balance. The analysis was performed for HTR-10 and PR-3000, two PBR-type reactors with different reactor core sizes. The results conclude that the larger bed as in PR-3000 is undergoing a smaller effect caused by the wall effect than in the smaller bed, such as the HTR-10. Comparison of three porosity models, Vortmeyer-Schuster, Mueller, and Benenati-Borislow, to the wall channeling, were also taken into account in the present study.

Partition method of wall friction and interfacial drag force model for horizontal two-phase flows

The improvement of thermal-hydraulic analysis techniques is essential to ensure the safety and reliability of nuclear power plants. The one-dimensional two-fluid model has been adopted in state-of-the-art thermal-hydraulic system codes. Current constitutive equations used in the system codes reach a mature level. Some exceptions are the partition method of wall friction in the momentum equation of the two-fluid model and the interfacial drag force model for a horizontal two-phase flow. This study is focused on deriving the partition method of wall friction in the momentum equation of the two-fluid model and modeling the interfacial drag force model for a horizontal bubbly flow. The one-dimensional momentum equation in the two-fluid model is derived from the local momentum equation. The derived one-dimensional momentum equation demonstrates that total wall friction should be apportioned to gas and liquid phases based on the phasic volume fraction, which is the same as that used in the SPACE code. The constitutive equations for the interfacial drag force are also identified. Based on the assessments, the Rassame-Hibiki correlation, Hibiki-Ishii correlation, Ishii-Zuber correlation, and Rassame-Hibiki correlation are recommended for computing the distribution parameter, interfacial area concentration, drag coefficient, and relative velocity covariance of a horizontal bubbly flow, respectively.

Thermal Hydraulic Analysis on the Water Lead Lithium Cooled Blanket for CFETR

A new type of Water Lead Lithium Cooled (WLLC) blanket that adopts the modular design scheme, water cooling the structure components, liquid PbLi as breeder and coolant, and SiC as the thermal insulator between PbLi and structures is under development as a candidate blanket concept for the Chinese Fusion Engineering Test Reactor (CFETR). Based on a poloidal-radial slice model, thermal hydraulic analysis is performed for this blanket to validate the feasibility of design goals. Results show that the present design can achieve the outlet temperature in the range of 600–700 °C, with all the material temperatures safely below the upper limits. A series of sensitivity analyses are also carried out. It indicates that the thermal conductivity (TC) of SiC would have a significant influence on the temperature field, streamlines and pressure drop; that is, lower TC of SiC can maintain the temperature of PbLi at a high level, and induce an increased number of vortices in the liquid PbLi flow as well as a larger pressure drop. On this basis, the joint effects of the TC of SiC and inlet velocity on the performance of blanket thermal hydraulics are analyzed, then the so-called “attainable region” is proposed. Finally, optimization design studies are carried out by decreasing the width of the front channel. Comparison results show that the present design is the most reasonable.

Thermal-hydraulic performance analysis of printed circuit heat exchanger precooler in the Brayton cycle for supercritical CO2 waste heat recovery

The supercritical carbon dioxide (SCO2) Brayton cycle is the most promising system in ship exhaust gas waste heat recovery. A printed circuit heat exchanger (PCHE) is used in the SCO2 Brayton cycle due to its high efficiency and compactness. To study the thermal–hydraulic characteristics of a PCHE as a precooler in a marine SCO2 Brayton cycle power generation system, a PCHE test platform was built, and substantial test results were obtained. These results suggest that the traditional heat transfer correlation does not accurately predict the heat transfer characteristics of SCO2. Subsequently, a simulation model established by a computational fluid dynamics (CFD) method is verified according to the experimental results, simulation calculations are carried out based on the design condition of the precooler in the marine Brayton cycle power generation system, and the axial, radial and global hydraulic and thermal characteristics of PCHE are analyzed. Finally, according to the CFD results, a new heat transfer correlation of SCO2 is obtained, thus providing a theoretical foundation for the design of precoolers in actual marine Brayton cycle power generation systems.

Study on the DLOFC accident of the GEMINI+ conceptual design of HTGR reactor with MELCOR and SPECTRA

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI Plus project. Behavior of the HTGR reactor under severe accident conditions was investigated and the maximum fuel temperature was observed. Due to application of the TRISO particles and SiC layers in the fuel element, no damage of the fuel is expected up to 1600°C. Under the cooperation in the project between Nuclear Research Group (NRG) and National Centre for Nuclear Research (NCBJ) a code-to-code calculations were carried out between the SPECTRA and MELCOR codes. SPECTRA code, developed by the NRG is a thermal hydraulic analysis code and MELCOR 2.1.6342 used by NCBJ developed by SANDIA National Laboratory is fast running severe accident code. Both codes have already HTGR specific models build in. The following accident was analyzed and will be presented: Depressurized Loss of Forced Circulation (DLOFC) with 65 mm break at the top of reactor vessel. The scenario was calculated applying following sets of assumptions: best estimate and conservative. Plant behavior was analyzed including primary and secondary side of the reactor. As the results of applying conservative assumptions, it was found that fuel temperature excides the acceptable limit of 1620°C. Therefore, changes in the core design were proposed by project participants. Analyses of the new core showed acceptable temperatures. In the paper the results of code-to-code comparison are presented. Both codes have shown a good agreement of presented following characteristics on maximum fuel temperature, relative power and Reactor Cavity Cooling System power, primary pressure and break flow.

Neutronics, thermal-hydraulics, and multi-physics benchmark models for a generic pebble-bed fluoride-salt-cooled high temperature reactor (FHR)

• A generic FHR design benchmark data set. • Neutronic, Thermal-Hydraulic and Pebble flow analysis. • Coupling of Monte Carlo, Porous Media and Discrete Element Modeling. • Multi-pass fuel cycle analysis with Monte Carlo based depletion calculation.

A distributed heat transfer model for thermal-hydraulic analyses in sewer networks

Thermal-hydraulic considerations in urban drainage networks are essential to utilise available heat capacities from waste- and stormwater. However, available models are either too detailed or too coarse; fully coupled thermal-hydrodynamic modelling tools are lacking. To predict efficiently water-energy dynamics across an entire urban drainage network, we suggest the SWMM-HEAT model, which extends the EPA-StormWater Management Model with a heat-balance component. This enables conducting more advanced thermal-hydrodynamic simulation at full network scale than currently possible. We demonstrate the usefulness of the approach by predicting temperature dynamics in two independent real-world cases under dry weather conditions. We furthermore screen the sensitivity of the model parameters to guide the choice of suitable parameters in future studies. Comparison with measurements suggest that the model predicts temperature dynamics adequately, with RSR values ranging between 0.71 and 1.1. The results of our study show that modelled in-sewer wastewater temperatures are particularly sensitive to soil and headspace temperature, and headspace humidity. Simulation runs are generally fast; a five-day period simulation at high temporal resolution of a network with 415 nodes during dry weather was completed in a few minutes. Future work should assess the performance of the model for different applications and perform a more comprehensive sensitivity analysis under more scenarios. To facilitate the efficient estimation of available heat budgets in sewer networks and the integration into urban planning, the SWMM-HEAT code is made publicly available.

Thermal-hydraulic analysis of the CFETR TF coils when subject to nuclear heat load

China Fusion Engineering Test Reactor (CFETR) has received much attention over the past several years, aiming at bridging the gap between the International Thermonuclear Experimental Reactor (ITER) and the Demonstration Fusion Reactor (DEMO). The toroidal field (TF) coils play an important role in the tokamak, which provide the main magnetic field to confine the plasma. In order to evaluate the feasibility of superconducting magnets used in CFETR, it is important to predict the magnet performance in terms of temperature margin during normal operation conditions. The simulations confirm the need to increase the mass flow rate, or decrease the hydraulic length of high field windings. The results show that the proposed reduction of hydraulic length is more effective to increase the minimum temperature margin.

Experimental performance analysis of a parabolic trough solar air collector with helical-screw tape insert: A comparative study

Heat transfer improvement efforts in parabolic trough air collectors (PTAC) are long-sought and much anticipated by many researchers, however, alternative heat transfer improvement methods such as helical screw tape insert inside the receiver tube has yet to be experimentally investigated. Therefore, in this study, a dual-function PTAC for low-medium temperature applications was experimentally tested in an indoor environment to assess its thermal characteristics. To increase the heat transfer performance, a helical screw tape was inserted along the pathway of air flow inside the receiver tube. To observe the effect, the PTAC system was tested with a helical screw tape insert and smooth absorber tube and the results were compared. Moreover, energy, exergy and thermal–hydraulic analysis of the PTAC was performed. Results show that the helical screw tape inserted PTAC always yields a much better performance in terms of overall collector efficiency and exergy efficiency. Overall results prove that the temperature rise improvement across the PTAC was remarkable, improving the efficiency by 56%–71% so that the role of the helical screw tape insert is to ensure an improved heat exchange. The findings validate the potential and elucidate specific conclusions for the practice of such heat exchange improvement methods in parabolic trough air collectors.

Coupled MELCOR/COCOMO analysis on quench of ex-vessel debris beds

The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-term coolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bed formed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of the MELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debris bed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integral analysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debris bed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on the bed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns. Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity pool saturation and containment venting.

Thermal-hydraulic analysis of the DTT CS and PF pulsed coil performance during AC operation

The EU DEMO fusion reactor, now entering in its conceptual design phase, will have to tackle several challenges, such as the long plasma pulse duration and the exhaust of very high heat fluxes from the plasma. The Divertor Tokamak Test (DTT) facility, a compact superconducting tokamak under construction at ENEA Frascati, will address the power exhaust problem in DEMO by testing several DEMO-relevant divertor solutions and operation scenarios.The DTT superconducting coils are the subject of the analyses presented here: as the tokamak must be very flexible to face different plasma scenarios, the design of the magnet system must be proven to be robust. In particular, the thermal-hydraulic performance of the coils operated in pulsed mode, namely the Central Solenoid (CS) and the Poloidal Field (PF) coils, will be analyzed. The CS is composed of 6 modules, layer-wound with Nb3Sn Cable-in-Conduit Conductors (CICCs), while the PF coils are pancake-wound. The two PF coils closest to the machine axis are wound using Nb3Sn CICCs, while the others adopt NbTi CICCs. All the coils are cooled by supercritical He (SHe) in forced flow.The pulsed operation of these coils induces AC losses, eroding their temperature margin: a detailed thermal-hydraulic model of the DTT pulsed coils is developed here using the 4C code. The model is then applied to the simulation of the two reference plasma scenarios, a single null and a double null, computing the minimum temperature margin. Different cooling options (single- or double-pancake) are investigated for the PF coils, while the sensitivity to the coupling time constant value is assessed for the CS.