Thermal-hydraulic Analysis Research Articles

Effect analysis of oxide layer on thermal hydraulic characteristics of LBE cooled reactor

For Lead-Bismuth Eutectic (LBE) alloy cooled reactor, oxidation corrosion is very important and would affect the thermal hydraulic characteristics of the reactor. In this paper, based on LETHAC code developed by Xi’an Jiaotong University (XJTU) for system level thermal hydraulic analysis LBE reactor, a simulation module for oxidation corrosion is developed and added. The influence of oxidation layer on the heat transfer characteristics of the reactor is conducted. Based on LESMOR reactor designed by XJTU, a simulation calculation is carried out under the long-term operation conditions of the reactor considering the growth of oxidation layer. The distribution of the oxide layer in the reactor, the change of temperature appreciation caused by the oxide layer over time and the distribution in the reactor are obtained. The presence of oxide layer will affect the fluid temperature and wall temperature of reactor. The longer the run time, the greater the impact. This paper could provide solid analysis methodology reference for the design of LBE reactor.

Development and application in multiscale and multiphysics methodologies in Spain: Present and future trends

In the field of reactor physics analysis, the coupling of multiple physics phenomena plays a crucial role in enhancing the accuracy of predictions and understanding complex systems. Multiphysics refers to the study and analysis of physical phenomena that involve multiple interconnected physical processes occurring simultaneously in a nuclear system to ensure essential aspects of reactor design, safety analysis, and optimization, among others. Multiphysics encompasses various domains of physics, such as Thermal-Hydraulics, Neutronics, Structural Mechanics, Radiation Transport, Chemistry, and Materials Science. On the other hand, the Multiscale approach involves combining high-fidelity models with lower-fidelity ones to accurately capture the relevant physical phenomena in a Nuclear Power Plant (NPP). Multiscale involves three main scales: system level, core level, and detailed analysis scale. This paper presents an overview of the main research performed by Spanish groups in the field of multiphysics and multiscale calculations, particularly on thermal–hydraulic analysis. This revision is focused mainly on two key aspects: thermal–hydraulic multiscale coupling and thermal–hydraulic / neutronic coupling. The paper also includes the expected future trends in this field.

Implementation of Solar Salt Properties Into ARIANT and Simulation of Pressurized Loss of Forced Circulation in a High Temperature Gas-Cooled Small Modular Reactor

Abstract Small modular reactors (SMRs) are actively being considered for use in Canada. Some proposed SMRs can make use of solar salt as an intermediate coolant for a heat storage system. The development of thermalhydraulic simulation tools is one of the key capabilities needed to examine the performance of SMRs and license this class of reactors. This article summarizes the implementation of molten solar salt fluid properties into the algorithm for analysis of network thermalhydraulics (ARIANT) and uses this code to simulate a high temperature gas-cooled SMR with helium and solar salt as its primary and secondary coolants during a pressurized loss of forced circulation (PLOFC) event. This work demonstrates the ability of ARIANT to simulate transient events in a two-loop reactor system consisting of helium and solar salt as coolants and helps to establish ARIANT as a tool for SMR analysis.

A methodology to quantify effects of constitutive equations on safety analysis using integral effect test data

To improve the predictive capability of a nuclear thermal hydraulic safety analysis code by developing a better constitutive equation for individual phenomenon has been the general research direction until now. This paper proposes a new method to directly use complex experimental data obtained from integral effect test (IET) to improve constitutive models holistically and simultaneously. The method relies on the sensitivity of a simulation result of IET data to the multiple constitutive equations utilized during the simulation, and the sensitivity of individual model determines the direction of modification for the constitutive model. To develop a robust and generalized method, a clustering algorithm using an artificial neural network, sample space size determination using non-parametric statistics, and sampling method of Latin hypercube sampling are used in a combined manner. The value of the proposed methodology is demonstrated by applying the method to the ATLAS DSP-05 IET experiment. A sensitivity of each observation parameter to the constitutive models is analyzed. The new methodology suggested in the study can be used to improve the code prediction results of complex IET data by identifying the direction for constitutive equations to be modified.

On the Use of the Jacobian-Free Newton Krylov Method to Generate One-Group Discontinuity and Super Homogenization Factors for Full-Core Neutron Diffusion Simulations

Coupled Monte Carlo (MC) and thermal-hydraulic analysis is valuable as a design or reference tool but can be slow, especially when implemented in a Picard iteration. Previous work has developed a novel prediction block to achieve convergence with fewer MC simulations. The prediction works in two stages: (1) a surrogate-like model predicts macroscopic cross sections on the fly and (2) a reduced-order neutronic model predicts the flux response to the updated cross sections. The main challenge with the prediction block is that the reduced-order neutronic model cannot reproduce the spatial flux distribution with high fidelity. This paper investigates the well-established Jacobian-free Newton Krylov (JFNK) method to preserve equivalency between a homogeneous (nodal diffusion) solution and a high-fidelity transport (MC) solution. Instead of performing multiple computationally consuming MC simulations, the nonlinear iterative approach iterates on correction parameters, e.g., assembly discontinuity factors (ADFs) or super homogenization (SPH) factors, using unexpensive nodal solutions. The JFNK approach does not require additional overhead from the MC solver to generate flux tallies. Further, the approach iterates on diffusion solutions produced directly from a desired code, thus ensuring that the parameters are compatible with that code. The approach is applied to correct a nodal diffusion solution for a realistic three-dimensional pressurized water reactor core. The results obtained in the paper show that the method is very successful in reproducing the heterogeneous solution (up to 2.5% difference in assembly flux for SPH and 0.3% for ADFs) without needing to modify the source code of the nodal diffusion solver. In addition, the results show that ADFs yield the best agreement and are also stable (i.e., weakly varying) when thermal-hydraulic fields are perturbed.

New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation

Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the Reactor Monte Carlo (RMC) code, offering multiple modes with varying levels of fidelity and computational requirements. The most precise mode is utilized in coupling simulations between RMC and the subchannel thermal-hydraulic analysis code SUBCHAN, incorporating an explicit moderator heating fraction in the coupling interface. The new treatment is verified against references from MCNP, Serpent, and OpenMC for three light water reactor (LWR) assembly cases, and great agreement is achieved. Energy deposition in different materials and components is emphasized in Kilowatt Reactor Using Stirling TechnologY (KRUSTY) modeling, and the results obtained using different modes are compared. The RMC-SUBCHAN coupling calculations for the three LWR assembly cases, employing the most accurate model, reveal a maximum increase of 94.6 K in the control rod centerline temperature, with a normalized energy deposition of 35.9% in the control rod regions. In the assembly case with gadolinium (Gd) burnable poison, a temperature increase of 7.3 K is observed in the Gd rod centerline, while the coolant outlet temperature decreases by 1.6 K due to the reduced explicit moderator heating fraction of 2.1%, compared to the constant 2.6% in the previous coupling scheme.

Validation of transient SYSTEM analysis code GINKGO based on a SLB TEST

Transient System analysis code is an important tool for reactor thermal hydraulic analysis. GINKGO is a transient system analysis in-house code developed by China General Nuclear Power Corporation (CGN). The code is primarily used for the analysis of Non-Loss of Coolant Accident (Non-LOCA)transients. Steam Line Break (SLB) accident is one of the representative accident condition of Non-LOCA transients. To validate the capability of GINKGO to simulate SLB accident scenario, a test of small non-isolatable main steam line break (10 %) at hot standby conditions, PKL III B5.1 test is selected. The characteristic of PKL facility, the PKL III B5.1 test sequence and the validation process of PKL III B5.1 test by GINKGO code are briefly introduced in the paper. In validation process, GINKGO simulates the steady state first and then simulates the transient process by setting reasonable initial conditions and boundary conditions. Then, the calculation results of key parameters including temperature in hot leg, temperature in cold leg, Steam Generator (SG) secondary side pressure and Pressurizer (PZR) pressure are compared with the experimental results. Through the rationality of the calculation results, it shows that GINKGO has the capability to simulate the key phenomena in a typical SLB accident scenario.

Thermal-Hydraulic Analysis of Pebble Bed High Temperature Gas-Cooled Reactor After Shutdown Based on FLUENT Software

In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media model and user-defined scalar (UDS) function of FLUENT software. Then, the model was applied to the numerical simulation of the shutdown test of the 10 MW high temperature gas-cooled test reactor (HTR-10) at 9 MW power level, and the temperature distribution and flow field distribution in the reactor were obtained and compared with the results of the experimental data. The reliability of the model in this paper was verified. Based on the model, the effects of the water-cooled panel temperature and the initial core temperature on the thermal-hydraulic characteristics of HTR-10 after shutdown were further explored. The results show that there is a decoupling phenomenon between the residual heat transfer within the core and the heat dissipation of the pressure vessel. The initial core temperature has relatively little effect on the heat dissipation and maximum temperature of the pressure vessel, but it has a significant impact on the thermal characteristics of the core area.

Development of transient thermal-hydraulic analysis model for the two-phase loop thermosyphon

To comprehensively and accurately simulate and predict the flow and heat transfer characteristics of the two-phase loop thermosyphon (TPLT), a two-fluid two-phase model was developed. A six-equation set was formulated to account for two-phase flow, encompassing mass equations, momentum equations, and energy equations. The interactions between the fluids in terms of mass, momentum, and energy were also taken into account. To obtain numerical solutions for the partial differential equations within a staggered mesh framework, we employed the 2nd-order Lax-Wendroff scheme to obtain the discrete linear system, complemented by the inclusion of a flux limiter. The Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to address the solution of the linear equation system. Data from a small-scale TPLT test was used to assess the accuracy and reliability of both the physical models and the numerical method proposed in this study. Simulations were performed under two working conditions of the thermosyphon with filling ratio of 45 % and 64 %. The simulation results exhibited good agreement with the experimental data, demonstrating the accuracy of the model. Furthermore, the model is capable of providing information on the complex two-phase flow inside the TPLT that cannot be measured experimentally.

Development and preliminary verification of a 1D–3D coupled flow and heat transfer model of OTSG

Introduction: A simulation model was developed by coupling a one-dimensional (1D) system code and 3D CFD software, to analyze the three-dimensional (3D) flow and heat transfer characteristics of the once-through steam generator (OTSG).Methods: The shell side of the OTSG was simulated by FLUENT, and the tube side was simulated by the system code LOCUST. Through spatial mapping, the 1D and 3D simulations were coupled along the outer wall of the OTSG's helically coiled tubes.Results and Discussion: This coupling method enabled the acquisition of high-resolution flow and heat transfer characteristics of the OTSG, and the error of heat flux calculation result by the coupled model is within 15%. Through coupling simulation analysis of the prototype OTSG, it was found that the inlet and outlet temperature difference reached as high as 150°C. The unevenness of the radial temperature distribution increased along the flow direction, and the wake swing effect caused by the sweeping flow of the tube bundle at the exit position was evident. The results of this study provide reference and a coupled simulation method for the engineering design and thermal-hydraulic characteristics analysis of OTSG.

Conceptual design and supporting analysis of a double wall heat exchanger for an ARC-class fusion reactor Primary cooling system

In the fight against climate change through the progress of fusion technology, the Affordable, Robust and Compact (ARC) fusion reactor design is under development at Commonwealth Fusion Systems (CFS) and Massachusetts Institute of Technology (MIT)- Plasma Science and Fusion Center, in collaboration with the Ente Nazionale Idrocarburi (ENI) S.p.A. The reactor features demountable superconducting toroidal field coils and a replaceable vacuum vessel immersed in a Fluorine Lithium Beryllium (FLiBe) molten salt blanket. The low-pressure molten salt cools the divertors and the blanket, shields the magnets and acts as tritium breeder and tritium carrier. Tritium must be recovered efficiently from the salt to fuel the reactor, minimizing its inventory and its migration outside the FLiBe loop. The current work is aimed at presenting a preliminary design of a Double-Wall Heat eXchanger (DWHX) to be installed in an ARC-class reactor primary cooling system. The main feature is the presence of a flowing sweep gas gap which captures tritium and prevents its diffusion toward the secondary system. For the DWHX, thermal-hydraulic and tritium transport analyses were performed, considering a supercritical water Rankine power conversion system on the secondary side to carry out the calculations. To assess tritium extraction capabilities of the DWHX, the tritium extraction efficiency has been evaluated. Calculation results highlighted how the current configuration is not adequate to remove tritium, in fact the tritium extraction efficiency is very low. On the other side, it seems to act adequately as tritium permeation barrier.

Thermal-hydraulic performance analysis of the supercritical fuel printed circuit heat exchanger with zigzag flow channels

Thermal-hydraulic performance analysis of the supercritical fuel printed circuit heat exchanger with zigzag flow channels

Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

Development of thermo-physical property database and code of fusion materials for CFETR blankets

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap.

Two-group drift-flux model for dispersed gas-liquid flows in rod bundles

Interfacial transfer terms in gas-liquid two-phase flows are formulated as the product of interfacial area concentration (IAC) and flux. The interfacial area transport equation (IATE) is essential for obtaining IAC in transient and developing two-phase flows. Two-group IATE was developed to account for the difference in the interfacial drag force between two groups of bubbles, where spherical and distorted bubbles are categorized as group one, while cap, slug, and churn turbulent bubbles are categorized as group two. The rigorous two-group approach requires two-group gas momentum equations, resulting in one additional momentum equation in the two-fluid model. The mixture gas momentum equation is considered to avoid the additional momentum equation. The two-group drift-flux model is necessary to calculate the two-group gas velocity from the mixture velocity. The present study develops an approximation methodology to acquire subchannel average and rod bundle average two-group two-phase flow parameters based on the validated power law assumption and a limited number of local data. The two-group model is developed for the distribution parameter and drift velocity for dispersed two-phase flows in rod bundles. The total performance of the newly developed two-group drift-flux model is evaluated by two-phase flow data in rod bundles. The results show that the developed two-group drift-flux model could well predict two-group gas velocities in rod bundles. This model is useful for subchannel thermal hydraulic analysis codes as a complement to the conventional one-group drift-flux model.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Neutronics and thermal hydraulics analysis on a Lead-bismuth-cooled fast reactor

Breeding Lead-base Economical Safe System-Demonstration BLESS-D is a lead-bismuth-cooled fast reactor proposed by China State Power Investment Corporation Research Institute. It features a pool-type configuration of pressure vessel with a thermal/electric power of 300/100 MW. Neutronics and thermal hydraulics analysis are performed, mainly including a sensitivity study of the fuel components, fuel enrichment, and control-assembly arrangement on the lifetime, and the temperature filed for the fuel assembly with the largest thermal power. The results show that compared with PWR, the 239Pu production for BLESS-D is significantly higher, and the quantity of MAs is also reduced, while the productions of LLFPs are at the same scale. It has also been found that under steady state conditions, the largest temperatures of the fuel pellet, cladding and coolant are all below relevant criteria, showing a good performance in the thermal hydraulics for the design. Optimizing the fuel loading scheme for achieving a safer performance would be a part of the future’s work.

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.

Estimating Allowable Operator Action Time for Mitigating Multiple‐Failure Accidents in Korean Advanced Power Reactor

In the case of multiple‐failure accidents in light water reactors, it is imperative to demonstrate that a significant fuel damage can be prevented via prearranged mitigation measures performed by operators. However, the time elapsed from the initiation of the event to the implementation of mitigation actions significantly influences the progress and consequences of the accident. Therefore, assuming a specific time for operator mitigation actions does not provide sufficient insights for safety assessments, thereby calling for further investigations into the effect of the time delay from the prescribed value. In this study, the allowable operator action time for multiple‐failure accidents of the Advanced Power Reactor 1400 was estimated by means of the sensitivity analysis using multi‐dimensional analysis for reactor safety‐KINS standard (MARS‐KS), a best‐estimate thermal–hydraulic analysis code used by the Korean nuclear regulatory authority. Three multiple‐failure accidents were investigated, including small‐break loss‐of‐coolant accident with the loss of safety injection (SI), intermediate‐break loss‐of‐coolant accident with the loss of SI, and total loss of feedwater accident. The upper limit of the mitigation action time, which prevents core damage, was estimated based on the requirement for peak cladding temperature. This was in turn used to derive the tolerable delay from the assumed action time in the Korean accident management program for each multiple‐failure accident.

Numerical analysis of reflood heat transfer and large-break LOCA including CRUD layer thermal effects

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.