Thermal-hydraulic System Code Research Articles

ARIANT Assessments for CANDU Analysis Using RD-14M Experiments

ARIANT (AlgoRIthm for Analysis of Network Thermalhydraulics) is a Canadian Nuclear Laboratories system thermal-hydraulic code for the modeling and analysis of two-phase flow and heat transfer for pressurized heavy water reactors, light water pressurized water reactors, and advanced reactor applications. This paper presents ARIANT models and simulations of RD-14M experiments, including small-break loss-of-coolant accidents, large-break loss-of-coolant accidents, loss-of-flow accidents, station blackout, and natural circulation, that are representative of accident scenarios in a CANDU reactor. ARIANT predictions of pressures, flow rates, temperatures, and void fractions are compared against the steady-state and transient data over the course of the tests. The results show that ARIANT predicted the key parameters with reasonable accuracy, as well as the overall behavior of the five transient events. These assessments support ARIANT’s applicability to the corresponding CANDU design-basis accidents and demonstrate ARIANT as an alternative to existing system thermal-hydraulic codes for CANDU safety analysis.

Implementation of a Drift Flux Model into SAM with Development of a Verification and Validation Test Suite for Modeling of Noncondensable Gas Mixtures

The advanced thermal-hydraulic system code, System Analysis Module (SAM), was originally developed for the modeling of single-phase flow in advanced reactors. It has since been expanded to include a four-equation drift flux model for the modeling of two-phase flows containing a noncondensable gas. The model was expanded to support the modeling of molten salt reactor (MSR) designs in which the fuel is directly dissolved in the circulating coolant. These designs have shown that circulating gas bubbles can play an important role in the management of fission products and the operational behavior of the reactor. A drift flux model was implemented to more accurately capture the localized behavior of the void in the core and its impact on the mass transfer of fission products. A thorough assessment of the new model was performed by developing a verification and validation test suite. Verification problems were designed to test all major terms in the new governing equations. The new model converged to the correct solution at the expected order of accuracy for all verification cases. The validation cases included a wide range of flow and void conditions in different pipe geometries. Although higher void experiments show a slight underprediction of void by the drift flux model, experiments that aim to reproduce Molten Salt Reactor Experiment (MSRE) experimental conditions show good agreement with the model. The gas transport model was activated for a SAM model of the MSRE to demonstrate that it can be used in a more complex model. This gas transport model will be used along with an interfacial area transport equation being implemented in SAM for the prediction of mass transport behavior in MSR conditions.

Effect of inclination on SB LOCA transient of a floating nuclear reactor in MARS-KS analysis using moving reactor model

As the imperative to reduce Green House Gas emissions grows within the maritime sector, nuclear power emerges as a promising solution. Especially, offshore floating nuclear power plants have garnered significant attention and development efforts from various institutions and nations. These types of reactors provide unique benefits such as extended refueling periods for advanced safety and exceptional mobility, facilitating access to remote regions. However, the external forces are induced by operating on a moving condition, that can impact the thermal–hydraulic characteristics of various hydraulic components in the entire reactor system. Addressing the safety issues of offshore floating reactors requires conducting thermal–hydraulic analyses under diverse oceanic motions. To achieve this objective, this study centered on analyzing accidents in an offshore floating reactor, specifically the BANDI-60, targeting the Small Break Loss of Coolant Accident induced by a double-ended guillotine break in the Direct Vessel Injection line. This study utilizes the system thermal–hydraulic code MARS-KS moving reactor model considering various tilting conditions and a BANDI-60 nodalization with a quarter-divided reactor pressure vessel and containment vessel volumes in the circumferential direction. The inclination conditions include static inclination along the x and y directions. Significantly, static inclination, particularly when the break aligned downward with the inclination direction, emerged as the most critical condition affecting reactor core integrity. Additionally, the containment vessel volume affects the equilibrium pressure after a break occurs, with smaller volume resulting in smaller peak cladding temperature values in static inclined conditions.

Validation of a Hybrid Domain Overlapping Coupling Between SAM and CFD Against the TALL-3D Transients

The system thermal-hydraulic (STH) code SAM has been coupled to the computational fluid dynamics (CFD) code Simcenter STARCCM+ utilizing a hybrid domain overlapping coupling method in space and an explicit coupling in time. The coupling aims to extend the STH code’s applicability to scenarios where local momentum and energy transfers are important yet difficult for STH standalone models to capture, such as three-dimensional (3D) mixing and thermal stratification. The coupling method’s numerical stability was previously verified against multiple test cases including two closed-loop configurations, and it was validated against a double T-junction experiment with 3D scalar mixing. In the present work, the method is validated against the TALL-3D STH/CFD coupling benchmark facility. TALL-3D is a liquid-metal facility with a large, pool-type enclosure (test section) that exhibits 3D flow effects to be modeled by a CFD code. The rest of the system exhibits approximately one-dimensional behavior that is well predicted by a STH code. First, a STAR-CCM+ CFD model of the 3D test section is validated against experimental temperature data. Then a SAM-STARCCM+ coupled model is validated against six different TALL-3D steady states and includes SAM standalone results for comparison. Last, the coupled model is validated against two TALL-3D transients: one exhibiting flow reversal in the test section and one exhibiting nonlinear, limit cycle oscillations (LCOs). For the first transient, the SAM-STARCCM+ coupled model properly predicts an increase in the test section’s inlet temperature during flow reversal, and this increase is not predicted by the SAM standalone model. The coupled model also better predicts initial flow recovery and subsequent oscillations as the system approaches a final natural circulation state. For the second transient, no true final steady state is observed due to LCOs. Neither the SAM-STARCCM+ coupled model nor the SAM standalone model can perfectly capture the experiment’s changing oscillation frequency during the transient. Nevertheless, the coupled model does reproduce the general oscillatory feedback observed. This is a significant achievement because the present coupling only uses an explicit coupling in time, as opposed to a tighter, semi-implicit coupling. In comparison, STH/CFD coupling efforts previously performed by other authors required a semi-implicit coupling to obtain similar results to this work. The strengths of the novel coupling scheme validated in the present paper lie in the superior convergence characteristics and the much simpler, nonintrusive implementation.

Analysis of AP1000 Small-Break Loss-of-Coolant Accident Using Reactor Transient Simulator

The Westinghouse Electric Company’s Advanced Passive Reactor (AP1000) is characterized by the incorporation of passive safety systems (PSSs) designed to ensure core cooling during transient events. The assessment of PSSs requires evaluation of their performance through a combination of experiments and simulations employing various thermal-hydraulic codes. In addition, detailed evaluation of PSSs for a specific reactor system transient analysis such as loss-of-coolant-accident analysis supports understanding representative integral effects test facility development and the further evolution model development and assessment process. Developing a reactor system code is a complex and time-consuming process that requires significant engineering expertise and effort. It can take several months to even years to complete in the early stages of reactor system design and analysis. However, this process can be expedited through the use of transient simulator models for similar reactor systems, which can be used for lesson learning and training purposes. This study uses the Personal Computer Transient Analyzer (PCTRAN) code. The main advantage of PCTRAN is its ease of use and ability to run faster than real time. This study presents the results obtained for a small-break loss-of-coolant accident (SBLOCA) for two breaks using the full version (licensed) of PCTRAN. The purpose of this investigation is to evaluate the overall system behavior during the postulated SBLOCA event as well as assess the capability of the PCTRAN code to reproduce the system response during transient events. The obtained results were compared with the Westinghouse NOTRUMP system code. The PCTRAN code proved to be reliable in predicting the qualitative behavior of the system in both transient cases. As for the system response, it was found that it is contingent on the activation time of the PSSs. The differences in reactor coolant system pressure between the two codes were attributed to the critical flow model and simplification of mass and energy balance. Despite PCTRAN’s limitations, it can still provide a reasonable prediction of various reactor parameters such as pressure, mass flow rate, and void fraction during a SBLOCA scenario. It is worth noting that PCTRAN currently employs a bulk approach similar to that of the Modular Accident Analysis Program (MAAP) and MELCOR codes. However, the upcoming version of PCTRAN will include an artificial intelligence–based detection and accident prevention system, as well as different models for different reactor components. Consequently, PCTRAN has the potential to be upgraded to match the system thermal-hydraulic codes of the U.S. Nuclear Regulatory Commission and become more widely used in cybersecurity to safeguard nuclear power plants from cyberattacks.

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

Transient simulation and analysis of a supercritical CO2 heat removal system under different abnormal operation conditions

The supercritical carbon dioxide (sCO2) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible.

Comparison of thermal hydraulic performance between horizontal and vertical steam generators in nuclear power plants

Abstract Steam generators (SG) for Nuclear Power Plants (NPPs) are important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. Vertical steam generators are used normally in North America, Europe and South East Asia, but horizontal steam generators are currently of great importance to reduce building heights and construction costs that govern the overall cost of new nuclear power plants. The objective of the current study is to measure the performance of nuclear steam generators in NPPs through comparative assessments of vertical and horizontal configurations. For this purpose, the vertical U-tube steam generators of a typical 4-loop Pressurized Water Reactor (PWR) was replaced by a VVER-1000 horizontal steam generators to evaluate the horizontal SG performance, at the same plant conditions and characteristics, in comparison with the vertical SG. The thermal hydraulic system code RELAP5 was used in the current work to perform the required analysis, during steady and transient scenarios.

Opposing Mixed Convection Heat Transfer for Turbulent Single-Phase Flows

Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. The objectives of this study are to gain a comprehensive understanding of opposing flow mixed convection heat transfer and to establish the prediction methodology by evaluating existing correlations and models. Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Because the definition of nondimensional parameters and their validated range confirmed by experiments differ for each correlation reported in previous studies, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing flow mixed convection and the single-phase heat transfer correlations implemented in the thermal–hydraulic system codes. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The effect of the difference in the thermal boundary conditions, i.e., uniform heat flux and uniform wall temperature, on the turbulent mixed convection heat transfer coefficient is not substantial. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids. The authors investigated the extrapolation performance of the mixed convection heat transfer correlations for a wide range of nondimensional parameters and observed that the Jackson and Fewster, Churchill, and Aicher and Martin correlations exhibit excellent extrapolation performance with respect to natural and forced convection flows, indicating that they can be applied beyond the parameter range validated experimentally.

Thermal-hydraulic study of the EU-DEMO Helium Cooled Pebble Bed Breeding Blanket Primary Heat Transport System

The European DEMO reactor will work under normal operating conditions in accordance with a pulsed duty cycle. However unplanned and planned transients of plasma overpower may occur compromising the integrity of its plasma-facing components structures. Consequently, adopting appropriate tools is essential to accurately and consistently model the thermal-hydraulic response of the involved cooling systems under both normal operating conditions and accidental events. Given this background, the University of Palermo, in collaboration with EUROfusion, started a research work to study the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) of the DEMO device under steady-state and transient conditions. The activity has been performed adopting a computational approach, employing the thermal-hydraulic system code TRACE version 5.0 patch 6. The key point of the work has been the code-to-code benchmark with the outcomes previously obtained with the RELAP5-3D code, to estimate the impact of the physical models, numerical resolution schemes and modelling techniques adopted on the predictive capabilities of the system codes considered. The models and the analysis results are presented and critically discussed herein.

Preliminary thermal-hydraulic deterministic safety analysis of an in-vessel LOCA for the DTT facility

The DTT (Divertor Tokamak Test) facility is a new experimental tokamak under development by an Italian consortium in cooperation with several high-standard European laboratories. The ENEA division FSN-SICNUC is involved in the project for carrying out deterministic safety analysis of postulated accidents. In the first year of activity, it has been analyzed an In-Vessel LOCA (IVLOCA) scenario. The IVLOCA scenario selected is caused by a break in the divertor’s cassette cooling tubes with the consequent release of coolant inside the Vacuum Vessel (VV). The best estimate thermal-hydraulic system code TRACE, developed by USNRC, was selected to conduct the preliminary thermal-hydraulic analysis of this accident. DTT is currently under design, so not all the data are frozen. Therefore, based on some engineering assumptions and scaling considerations from similar facilities, the nodalization of DTT VV was developed using the three-dimensional TRACE component “VESSEL”. Then, a sensitivity analysis was carried out to simulate the break and the consequent water injection in the VV. This was done to compare the system behavior and test different nodalization approaches. Subsequently, the data of the DTT divertor cooling system were used to run additional simulations. The results allow to compare different nodalization approaches and to have a preliminary estimate of the pressure peak and temperature behavior in the VV for an IVLOCA. Finally, a first uncertainty analysis was carried out using the DAKOTA toolkit, coupled with TRACE code in SNAP. Two uncertain input parameters were selected: the rupture area of a cooling divertor tube and the temperature of the divertor coolant. The uncertainty analysis allows having a wider spectrum of system behavior and a preliminary insight on the dispersion of the VV pressure, selected as figure of merit. This paper aims to show the results of this preliminary analysis, characterizing the phenomena that occurred during the selected transient.

BEPU analysis of an Upgraded ICE facility test by TRACE/DAKOTA coupling

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the safety issues in fusion nuclear plants. The best estimate thermal-hydraulic system codes adopted to perform deterministic safety analysis should be validated against the phenomena typical of accidental transients in fusion installations. TRACE (TRAC/RELAP Advanced Computational Engine), best estimate thermal-hydraulic system code developed by USNRC, has been adopted to simulate an ICE. The calculated results have been compared to the experimental data obtained in one test performed in the upgraded Integrated ICE facility at JAERI. In this updated configuration the pressure suppression system is connected to the top of the plasma chamber instead of the bottom of the vacuum vessel. The facility nodalization has been developed in the SNAP environment/architecture. To qualify the code and the nodalization, an accuracy evaluation has been performed both from a qualitative and quantitative point of view. Then, considering the presence of some uncertainties in the input-deck development, an uncertainty analysis has been carried out. The probabilistic method to propagate the input uncertainties has been selected and the analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in SNAP. In the uncertainty analysis, some relevant statistical parameters have been considered to characterize the dispersion of the results and the correlation between the uncertain input parameters selected and the PC pressure chosen as figure of merit.

Analysis of main steam line break accident on a BWR test facility using TRACE

PUMA-E is an integral test facility, which is used for extensive experimental investigations to study the behavior of Boiling Water Reactor (BWR) plants under accident conditions. The TRACE code is the best estimate thermal hydraulic system code developed by the U.S. Nuclear Regulatory Commission for performing nuclear power plant safety analysis. The aim of the present work is to conduct a short validation study of the TRACE code applied to the PUMA-E facility. Firstly, a TRACE model of the PUMA-E facility was developed based on the existing RELAP5 input deck. Then, an additional facility TRACE model with reactor pressure vessel modeled by three dimensional “VESSEL” component was also established. After that, a Main Steam Line Break accident test was simulated, and the calculation results were compared with experimental data and RELAP5 simulation results. The TRACE simulation capability to predict the accident phenomenon of BWR test facility was evaluated using both qualitative and quantitative methods. On the basis of the above research, the base case accompanying with one Gravity Driven Cooling System tank failure accident was carried out based on the three facility models. The validity and reliability of the simulation results of the different facility numerical models were mutually verified through a qualitative comparison of the simulation results.

Development of a Fully-Implicit Second-Order system Thermal-Hydraulics analysis code for reactor modeling and simulations

There has been continuous advancement in numerical techniques and software engineering in the field of system thermal-hydraulics over the past decade. The current state-of-the-art involves the use of fully-implicit high-resolution schemes in conjunction with the Jacobian-Free Newton-Krylov (JFNK) method for both single-phase and two-phase flows. However, the application of these advanced schemes in system thermal-hydraulics codes is not a common practice. Given the opportunity and the need to develop a new modern system thermal–hydraulic code from scratch, there is no doubt that these recent advancements should be leveraged. In this work, we introduce a new modern object-oriented system thermal-hydraulics code named RETA. RETA is characterized by several key features, including: second-order temporal and spatial discretization schemes for both incompressible and compressible flows, as well as single-phase and two-phase flows; fully-implicit solution schemes utilizing both the Newton and PJFNK methods; and an object-oriented design of major fluid components and physical models, which allows for easy extension. A series of verification and demonstration studies are conducted to verify the implementation, demonstrate the accuracy improvement, and evaluate the performance of the new code. It is concluded that results from the second-order scheme are generally more reliable. Without the concern of a code stability issue, the second-order scheme should be the preferred and default choice for a code user or an analyst.

SiRENE: A new generation of engineering simulator for real-time simulators at EDF

For Safety Assisted Engineering works, real-time simulators have emerged as a mandatory tool among all the key actors involved in the nuclear industry (utilities, designers and safety authorities). EDF, Electricité de France, as the leading worldwide nuclear power plant operator, has a crucial need for efficient and updated simulation tools for training, operating and safety analysis support.This paper will present the work performed at EDF/DT to develop a new generation of engineering simulator to fulfil these tasks. The project is called SiRENE, which is the acronym of Re-hosted Engineering Simulator in French. The project has been economically challenging. Therefore, to benefit from existing tools and experience, the SiRENE project combines:- A part of the process issued from the operating fleet training full-scope simulator.- An improvement of the simulator prediction reliability with the integration of High-Fidelity models, used in Safety Analysis. These High-Fidelity models address Nuclear Steam Supply System code, with CATHARE thermal-hydraulics system code and neutronics, with COCCINELLE code.- And taking advantage of the last generation and improvements of instructor station.The intensive and challenging uses of the new SiRENE engineering simulator are also discussed. The SiRENE simulator has to address different topics such as verification and validation of operating procedures, identification of safety paths, tests of I&C developments or modifications, tests on hydraulics system components (pump, valve etc.), support studies for Probabilistic Safety Analysis (PSA). etc. It also emerges that SiRENE simulator is a valuable tool for self-training of the newcomers in EDF nuclear engineering centers.As a modifiable tool and thanks to a skillful team managing the SiRENE project, specific and adapted modifications can be taken into account very quickly, in order to provide the best answers for our users' specific issues.Finally, the SiRENE simulator, and the associated configurations, has been distributed among the different engineering centers at EDF (DT in Lyon, DIPDE in Marseille and CNEPE in Tours). This distribution highlights a strong synergy and complementarity of the different engineering institutes at EDF, working together for a safer and a more profitable operating fleet.

Validation of modeling scheme implementing MULTID component of MARS-KS code to predict shell-side pressure drop in shell-and-tube type heat exchangers

A modeling scheme to predict the pressure drop on the shell side of the shell-and-tube type heat exchanger is developed, which is using the system thermal hydraulic code, MARS-KS. In the present scheme, the shell side of the heat exchanger is modeled by the MULTID component in rectangular shape in which, the part blocked by the segmental baffle plate and the opening part by the baffle cut are modeled by separate nodes. The form loss coefficients and its distribution for the junctions of the shell side are estimated by the assumption of the same pressure drop in each cross pass of the shell and then finalized by iterative calculation. To prove the validity of the modeling scheme, calculations are made to simulate the experiments conducted at the test heat exchanger having a size used in the industries. It is discussed whether each effect of the size of inlet/outlet nozzle and the height of the baffle cut on the overall pressure drop of the heat exchanger is properly predicted and consistent with the experimental observations. Also, the effectiveness of consideration of leakage at the clearance between the shell and the segmental baffle plates is discussed in resolving the problem of over-predicted pressure drops in the high flow region. For the simplification, the use of one-dimensional modeling to this problem is attempted. The applicability of the present method and results to the different heat exchangers is also discussed.

Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility

The safety injection (SI) system might be put into use, and the coolant at high temperature will mix with the supcooled water under Loss Of Coolant Accident (LOCA). The thermo-hydraulic phenomena associated with the thermal mixing of coolant and supcooled water will directly affect the judgment on core reflooding. China General Nuclear Power Group (CGN) independently developed and designed a thermal hydraulic system code named LOCUST, and the verification of thermal mixing models in LOCUST is necessary. This paper will introduce the thermal mixing tests at the T-junction based on Emergency Core Cooling System in Xi'an Jiaotong University (ECCS-XJTU) experimental facility, which were mainly conducted for the mixing between subcooled water injected from the SI pipe with a range of 25–125 kg/h in mass flow and the pure steam in the primary pipe with a range of 100–500 kg/h in mass flow. Besides, the fluid temperature and the dynamic vapor quality after mixing in tests are analyzed, and the simulations of 25 thermal mixing tests using the LOCUST 1.2 code are performed. The results show that the maximum relative error of LOCUST in mass flow of liquid is within 13.8 %, and the maximum relative error of LOCUST in temperature is within 8 %, which validates the reliability and accuracy of simulations of LOCUST for two-phase thermal mixing in LOCA.

Benchmark analysis of the FFTF LOSWOS test #13 with OpenMC and THACS

In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi’an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie’ outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR.

Detail CFD simulation of the integrated reactor pressure vessel of SMART

This paper presents a detailed analysis of the full Reactor Pressure Vessel (RPV) of SMART with ANSYS CFX 20.2. SMART is a water-cooled SMR designed by KAERI for an operation with forced convection. The thermal power of SMART amounts to 330 MWth. In the integrated SMR-concept, the main primary loop components such as the helical steam generators, pressurizer and canned pumps are located inside the RPV. The CFD model uses a concept of full geometrical detail resolution with partial boundary inflation of the mesh and porous media modelling with sources for momentum and energy for components like the core, pumps and steam generators. The detailed results obtained with a steady-state RANS simulation using ANSYS CFX 20.2 are very promising as the comparison of selected core parameters of the CFD-simulation with the ones of the system thermal hydraulic code TRACE have shown.

Development of a one-dimensional system code for the analysis of downward air-water two-phase flow in large vertical pipes

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as ⟨jg⟩/⟨j⟩ increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.