Steady-state Thermal-hydraulic Analysis Research Articles

CFD analysis of PHWR exposed core under postulated severe accident condition

A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core.

Comprehensive feasibility assessment of cycle length extension for CNPGS Unit-3

The core cycle length is crucial parameter for effective and economic operation of a nuclear power plant. Nuclear industry has great emphasis on improving the feasibility and sustainability of nuclear power generation. The cost of the electricity can be reduced by increasing the cycle length without violating the safety margins. Various methods and strategies have been examined to extend a Pressurized Water Reactor (PWR) core cycle length with optimized loading patterns. The objective of this work is to analyse the feasibility of the system to extend the cycle length of Chashma Nuclear Power Generating Station (CNPGS) unit-3, currently operating in Pakistan from 12 months to ∼18 months. To achieve the objective, fuel enrichment is increased from 3.4% to 4.1% with Gd2O3 as burnable poisons to suppress the excess reactivity. The multi-cycle analyses showed promising results and the core cycle length has been extended to ∼18-month for CNPGS unit-3 keeping power peaking factor, temperature coefficients and discharge burnup within design limits. The method can be confidently applied at any cycle of the power plant. Additionally, fuel and moderator temperatures at average and hot channels at beginning of life (BOL) and the end of life (EOL) are predicted using the 1-D steady state thermal hydraulic analysis. The peak to average temperatures at BOL and EOL are found well below the reference value (1.6875).

Validation of DRAGON5 and DONJON5 computer codes with CNPGS Unit-3 cycle-1 design calculations

The necessity to replace outdated analogue nuclear system calculations as well as the desire to enhance and guarantee adequate levels of plant availability and safety have led to an increase in the usage of software based design and simulations of nuclear power plants in recent years. In this study, the computer codes DRAGON5 and DONJON5 are validated using design information for 300MWe Pressurized Water Reactor (PWR). The codes are used to perform design analysis of Unit-3 Cycle-1 of the Chashma Nuclear Power Generation Station (CNPGS). The analysis includes calculations for criticality, radial power distribution, reactivity coefficients, and control rod worth under various core conditions. The results are validated against the measured data. The difference in the critical boron concentration (CBC) for Hot Zero Power (HZP) and Hot Full Power (HFP) at All Rod Out (ARO) conditions are 16 pcm and 20 pcm respectively. It has been observed that, the core radial power distribution is in good agreement with the design data with maximum relative errors of 6.2% and 1.5% at HZP and HFP conditions respectively. The validation confirms that, the codes can be employed with confidence for the design calculations of 300MWe PWRs (CNPGS). Furthermore, the 1-D steady state thermal hydraulic analysis is also performed to predict the temperatures of the fuel and coolant at both the average and hot channels at the beginning of life (BOL) and the end of life (EOL). It is found that the peak to average values of fuel centerline temperature at BOL for average and hot channels are 1.084 and 1.17 respectively, while at EOL for average and hot channels are 1.059 and 1.069 respectively, which are in good agreement with the reference peak to average value of 1.6875.

Reactor physics calculation and measurement during first approach to criticality of Apsara-U research reactor

Apsara-U is 2 MW swimming pool type research reactor with low enriched uranium fuelled core. On 10th September 2018, Apsara-U was made critical at 18:41 hrs. After the First Approach to Criticality (FAC) of Apsara-U, low power physics experiments were carried out. The experiments included reactivity worth measurement of all reactivity devices and measurement of the isothermal temperature coefficient of reactivity. Inflated reactivity worth of Shut off Rods as calculated from the power from some of the neutronic channels were obtained on account of shadowing effect of neutron reaching the detector locations. Further the isothermal coefficient of reactivity was observed positive. In this paper, we discuss the mentioned Reactor Physics measurement and related theoretical calculation during FAC that includes solving the benchmark problem for validating the calculation tools, and steady state thermal hydraulic analysis, Reactivity initiated transients for the FAC core configuration for ascertaining the safety of the reactor.

Feasibility of breed-and-burn reactor core design with nitride fuel and lead coolant

The global demand for energy is constantly increasing, and reliable, sustainable energy sources are urgently needed. Nuclear energy is one such source, and many thermal reactors are currently in operation. As a result, a huge amount of depleted uranium has been stored, and reprocessing plants or fast reactors operating in breed-and burn (B&B) mode are necessary for effective utilization of uranium source. In the present paper, we investigate the feasibility of a Rotational Fuel-shuffling breed-and-burn reactor with Nitride fuel and Lead coolant (RFBB-NL) with neutronic and thermal-hydraulic analyses. In order to establish a high neutron-economy B&B mode, the core is established with a Rotational Fuel-shuffling (RF) scheme. The present study confirms the feasibility of a small LFR design functioning in B&B operating mode by optimizing core height and operating parameters. We kept the maximum temperatures of fuel, cladding, and coolant under their constraints by performing a steady state thermal-hydraulic analysis for the hot channel of a shorter core with higher power.

An Intelligent Optimization Method for Preliminary Design of Lead-Bismuth Reactor Core Based on Kriging Surrogate Model

To meet the numerous application demands of lead-bismuth reactors, different design optimization tasks need to be conducted on these reactors based on the existing reactor core solutions. However, the design optimization of lead-bismuth reactors is a challenging task because it is a complex, multi-dimensional, and nonlinear constrained problem. To resolve these issues and improve the efficiency of design optimization, a new method, called the KSM-OLHS-SEUMRE method, based on the Kriging surrogate model (KSM), orthogonal Latin hypercube sampling (OLHS), and space exploration and unimodal region elimination (SEUMRE) algorithm is proposed in this study. Based on this method, a design optimization program of lead-bismuth reactors (DOPPLER-K) is developed, which realizes functions like sample point generation, optimization analysis, pre-post processing of reactor calculation, coupling of the Reactor Monte Carlo (RMC) calculation code and the Steady-state Thermal-hydraulic Analysis Code (STAC). Further, taking lead-bismuth reactors SPALLER-4 and URANUS as prototypes, the proposed intelligent optimization method for preliminary design of lead-bismuth reactor core is verified. The results show that this method can rapidly and accurately find the target scheme satisfying the optimization conditions, and it is three orders of magnitude faster than pure Monte Carlo calculation. Compared with the initial core scheme of URANUS, the optimization rates of fuel loading, total core mass, active zone volume, and total core volume are reduced by 10.8, 11.5, 18.1, and 17.1%, respectively. These results validate the feasibility and efficacy of the proposed method for design optimization of lead-bismuth reactor core.

Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor

The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.

Steady-State Thermal-Hydraulic Analysis of Steam Generator with Axial Economizer Based on Node Method

Based on the one-dimensional flow hypothesis, heat transfer hypothesis, and two-phase thermal equilibrium hypothesis, a lumped parameter method and a distributed parameter method are used to establish a one-dimensional steady-state thermal-hydraulic analysis model of the steam generator with axial economizer. Using C++ language programming, the calculation results are compared with the design values of the thermal-hydraulic parameters of a typical steam generator with axial economizer. The results show that the relative error of most of the overall parameter calculation results is within 3%, which verifies the rationality of the model. The analysis of the temperature, cavitation share, pressure and other parameters of the steam generator along the direction of the measured fluid flow trend consist well with the results of thermal hydraulics and qualitative mechanism analysis, indicating that the model and solution method established in this paper can accurately predict the steady-state thermal hydraulic parameter distribution of the steam generator with axial economizer.

Feasibility of Sodium-Cooled Breed-and-Burn Reactor with Rotational Fuel Shuffling

The objective of the present study is to show that it is feasible to establish the breed-and-burn (B&B) mode of operation with rotational fuel shuffling in the S-PRISM core based on neutronic and thermal-hydraulic analyses. The results quantified the impact of major core design choices on the criticality of a core that uses sodium as a coolant material and HT9 steel as structural material. The design variables examined include the binary metallic fuel U-Zr with different weight percentages of zirconium as well as different core heights and fuel rod pitch-to-diameter ratios (P/Ds) in the fuel assembly. We found that a core using the binary metallic fuel U-Zr with 2 wt% zirconium, with a core height of 200 cm, a P/D of 1.086, and a core power of 400 MW(thermal), could overcome some major design constraints. It was also found that with shuffling intervals of 1125 to 1250 days, the core with rotational fuel shuffling was critical in the equilibrium state, and the possible average discharged burnup was from 274.8 to 305.3 GWd/ton HM. Reactor characteristics such as neutron flux and power profile were almost stable during the equilibrium cycle. A steady-state thermal-hydraulic analysis was performed for the hottest channel in the core. It revealed that both the fuel and cladding maximum temperatures were less than the melting point of the fuel and the chemical interaction temperature of the HT-9, respectively. The mixed coolant outlet temperature was somewhat below the temperature usually observed in sodium-cooled fast reactors. Thus, it appears that the S-PRISM can be principally designed to have a B&B core.

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Heat transfer characteristics of redan structure in large-scale test facility STELLA-2

The construction of STELLA-2 facility is on-going to demonstrate the safety system of PGSFR and to provide comprehensive understanding of transient behavior under DBEs. Considering that most events are single-phase natural circulation flow with slow transient, STELLA-2 was designed with reduced-height of 1/5 length scale. The ratio of volume to surface area in the vessel can relatively increase resulting in excessive heat transfer. Therefore, a steady-state thermal-hydraulic analysis was performed and the effect of design change to reduce the heat transfer through redan was investigated. The heat transfer through single wall redan in STELLA-2 was 3% of the core power, comparable to 1% of the core power in PGSFR. By applying the insulated redan, about 70% of decrease effect was observed. The effect on transient behavior was also evaluated. The conclusion of this study was directly applied to the STELLA-2 design and the modified version is under construction.

Physics study of block/prismatic-type HTGR design option for the Indonesian Experimental Power Reactor (RDE)

In our previous works, we have investigated the pebble-bed type high-temperature gas-cooled reactor (HTGR) as a candidate design for the 10 MWth Indonesian Experimental Power Reactor (“Reaktor Daya Experimental,” RDE). The once-through-then-out (OTTO) and multipass fuelling schemes with both uranium and thorium fuel cycles were considered. The heavy metal loading per ball and fissile enrichment as free design parameters were surveyed to obtain the optimum fuel compositions (the smallest fissile requirement per unit energy generated) for an average burnup target of 80 MWd/tU. Steady-state thermal-hydraulic analyses were also conducted, and no critical challenge was found since the reactor power density is low. Geographically the RDE site and the country, in general, are located in the so-called Ring of Fire (Circum-Pacific Belt) which is vulnerable against frequent seismic events. The nature of a block- or prismatic-type HTGR does not allow large and random movement of its fuel elements during a large seismic event which makes this type of HTGR relatively safer for these regions. The design also allows control and safety rods (absorbers) to be located inside the core by providing holes in graphite blocks which ensure the sub-criticality during an abnormal event. In the present work, under the same RDE design requirements and constraints, the block/prismatic-type 10 MWth HTGR is studied based on the Japanese 30 MWth High-Temperature Engineering Test Reactor (HTTR). Although the design study emphasized on the optimal fuel composition in the reactor physics lattice phase, a preliminary result of the full core reactor physics analytical work is given.

Power upgrading of WWR-S research reactor using plate-type fuel elements part I: Steady-state thermal-hydraulic analysis (forced convection cooling mode)

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MWth. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MWth.

Temperature and heat flux calculations for the maximum power channel of the TRIGA IAN-R1 research reactor

With cooperation of the International Atomic Energy Agency (IAEA), thermal-hydraulic calculations were carried out for conversion of the IAN-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. To establish thermal-hydraulic calculation and analysis research in Colombia, this program was carried out and included training, acquisition of hardware, software and natural convection flow calculations for the TRIGA IAN-R1 research reactor operating at 100 kW. The purpose of the study is to validate the steady state thermal hydraulic analysis that has been carried out by means of the NATCON code. This paper presents the results of the maximum axial temperature distribution for fuel, clad, and coolant. In addition, the Bernath critical heat flux with pool water temperature as a parameter is presented.

Transitional cores and fuel cycle analyses in support of MIT reactor low enriched uranium fuel conversion

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states.

Coupling of Neutronics and Thermal-Hydraulic Codes for Simulation of the MNSR Reactor

This study aims to evaluate a simplified one-dimensional thermal-hydraulic module (THM) established by the DRAGON5/DONJON5 codes that allow a multiphysics study of the Syrian Miniature Neutron Source Research (MNSR) reactor both in steady-state and transient conditions. The purpose of this paper is therefore to describe the THM, fully integrated and implanted in DONJON5 to allow coupling with neutronic modules existing in the same code and to perform steady-state thermal-hydraulic and safety analyses of the reactor. Then we compare the results given by the THM with the results obtained by the Program for the Analysis of REactor Transients (PARET)/Argonne National Laboratory (ANL) thermal-hydraulic code. In order to validate our PARET/ANL and the THM in DONJON5, the fuel center temperature as a function of core power was calculated and compared with the corresponding values of the PARET code. Moreover, we have calculated the departure from nucleate boiling ratio. The comparison of the results of this study showed a good correlation between the values obtained with the THM and the thermal-hydraulic PARET/ANL code.

Performance Test of the Air-Cooled Finned-Tube Supercritical CO2 Sink Heat Exchanger

This technical paper presents results of an air-cooled supercritical CO2 (sCO2) finned-tube sink heat exchanger (HX) performance test comprising wide range of variable parameters (26–166 °C, 7–10 MPa, 0.1–0.32 kg/s). The measurement covered both supercritical and subcritical pressures including transition of pseudocritical region in the last stages of the sink HX. The test was performed in a newly built sCO2 experimental loop which was constructed within Sustainable Energy (SUSEN) project at Research Centre Rez (CVR). The experimental setup along with the boundary conditions are described in detail; hence, the gained data set can be used for benchmarking of system thermal hydraulic codes. Such benchmarking was performed on the open source Modelica-based code ClaRa. Both steady-state and transient thermal hydraulic analyses were performed using the simulation environment DYMOLA 2018 on a state of the art PC. The results of calculated averaged overall heat transfer coefficients (using Gnielinski correlation for sCO2 and IPPE or VDI for the air) and experimentally determined values shows reasonably low error of + 25% and – 10%. Hence, using the correlations for the estimation of the heat transfer in the sink HX with a similar design and similar conditions gives a fair error and thus is recommended.

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

Effect Of U-9mo/Al Fuel Densities On Neutronic And Steady State Thermal Hydraulic Parameters Of Mtr Type Research Reactor

<p>The objectives of this research work are to carry out a detailed neutronic and steady state thermal hydraulics analysis for a MTR research reactor fuelled with the low enrichment U-9Mo/Al dispersion fuels of various uranium densities. The high density uranium fuel will increase the cycle length of the reactor operation and the heat flux in the reactor core. The increasing heat flux at the fuel will causing increase the temperature of the fuel and cladding so that the coolant velocity has to be increased. However, the coolant velocity in the fuel element has a limit value due to the thermal hydraulic stability considerations in the core. Therefore, the neutronic and the steady state thermal hydraulic analysis are important in the design and operation of nuclear reactor safety. The calculations were performed using WIMS-D5 and MTRDYN codes. The WIMS-D5 code used for generating the group constants of all core materials as well as the neutronic and steady state thermal hydraulic parameters were determined by using the MTRDYN code. The calculation results showed that the excess reactivity increases as the uranium density increases since the mass of fuel in the reactor core is increased. Using the critical velocity concept, the maximum coolant velocity at fuel channel is 11.497 m/s. The maximum temperatures of the coolant, cladding and fuel meat with the uranium density of 3,66 g/cc are 70.85°C, 150.79°C and 153.24°C, respectively. The maximum temperatures are fulfilled the design limit so reactor has a safe operation at the nominal power.</p>

Helical coil thermal-hydraulic model for supercritical lead cooled fast reactor steam generators

This paper deals with the presentation of the developments of an in-house code created for the steady state thermal-hydraulic analysis of helical coil steam generators. These components are of particular interest in the nuclear field both for normal and emergency operation because they are able to remove thermal power with a very high degree of compactness. This quality is particularly attractive for small modular reactors where constraints related to the volume of components are a critical issue. The model is able to perform calculations to validate the design of these components, run fast sensitivity analyses and show the spatial behaviour of the most important parameters of the flow such as temperature, pressure, heat transfer coefficients and heat flux. In this paper the additional features of the model such as the possibility to simulate supercritical water and liquid metals are presented. The choices made for the selection of semi-empirical correlations available in literature for the estimation of the heat transfer coefficients in these conditions are described and justified. Finally, the predictive capability of the updated model is validated against the nominal data of the BREST reactor, which is a lead fast reactor that make use of supercritical water.