Thermal-hydraulic Analysis Code Research Articles

Application of BPNN algorithm in thermal-hydraulic analysis of unwrapped LFR core

Due to the complex structure of the LFR fuel rod bundle, refined CFD simulation consumes significant computing resources and time, making it difficult to meet the immediate needs of engineering design and research. This paper introduces a method that integrates the sub-channel level thermal-hydraulic analysis code CorTAF-LBE with the BPNN algorithm to analyze the thermal-hydraulic phenomena within the LFR core. The method has been validated by refined simulation results of the THEADES test section, which indicate that the error is within an acceptable range. The feasibility of its application in LFR fuel assemblies was also evaluated. The flow characteristics and temperature distribution in unwrapped LFR rod bundles are investigated. Due to differences in fuel distribution and flow area, the temperature in the central channel is much higher than that in the side and trapezoidal channels. Three accident conditions, including inlet blockage, center blockage, and outlet blockage, were simulated. The corresponding peak temperatures of the fuel cladding were 670 K, 767 K, and 860 K, respectively, representing an increase of approximately 170 K compared to the nominal conditions. The analysis results effectively demonstrate the 3D temperature distribution of the fluid in the rod bundles, which is significant for reactor design and safety analysis of LFRs.

Development of the thermal hydraulic analysis code, TASS/SMR-S, under multi-dimensional motion for a reactor under inclination and rolling conditions

The development of a thermal hydraulic analysis code that considers ship motion is important because it would adjust for additional motion effects that are not present in a land-based system analysis code. The TASS/SMR-S code, a one-dimensional system analysis code, was modified to predict the thermal hydraulic behavior of a nuclear reactor under the conditions of an ocean environment. To validate the improved TASS/SMR-S code, the calculation results were compared with the analytical solutions for a conceptual problem involving swaying, rolling, and compounded motions; experimental data were collected under inclination and rolling conditions. The modified TASS/SMR-S code exhibited good agreement with analytical solutions for the conceptual problem and experimental results obtained under ship motion conditions. In conclusion, the modified code was able to simulate thermal hydraulic behavior under ship motion conditions.

CorTAF-LBE: A full scale subchannel-level thermal-hydraulic characteristics analysis code for LBR core based on OpenFOAM

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation.

Development of modular dynamic code for steam power conversion system in sodium-cooled fast reactor

Sodium-cooled fast reactor (SFR) is one of the most promising Generation IV reactor types. To improve operational safety and economy, SFR requires accurate prediction of the steam power conversion system. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR is developed in this study where each equipment is packaged as a separate module with modular modeling method. JFNK is adopted to solve ODE in this code, which expand the universality of the code. The model is validated by comparing with design results from the Daya Bay nuclear power plant and the BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. After validation of the code, the steady-state analysis of the SFR steam power conversion circuit is carried out where the flow and heat transfer process of key equipment can be accurately simulated considering a complete steam power conversion cycle. This work provides reference value for the thermal-hydraulic characteristics of steam power conversion system in SFR.

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.

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.

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.

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.

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.

A subchannel analysis code SACOS-Na for sodium-cooled fast reactor

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm.For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes.

Parallelization and optimization for a pin-by-pin whole core thermal–hydraulic analysis code

With increasingly developing high-performance computing (HPC) technology, pin-by-pin thermal–hydraulic simulations are now feasible. The in-house subchannel analysis code KMC-FBc has been extended to perform pin-by-pin calculations. To expedite the whole-core thermal–hydraulic analysis process, various parallelization and optimization strategies are proposed and implemented at the single-assembly and multi-assembly levels, including acceleration methods for solving large sparse matrix equation systems, memory and file input/output (I/O) optimization, hybrid MPI/OpenMP-based whole-core acceleration method and so on. Parameters for evaluating the parallel performance, such as the running time, speedup, and parallel efficiency, are tested on an HPC cluster. Results show that the maximum acceleration ratio for the whole-core problem with a mesh size in the tens of millions can reach approximately 120, with a parallel efficiency of around 32%, which demonstrates the rationality and effectiveness of the parallel algorithm and optimization methods and lays the foundation for future high-resolution and high-fidelity thermal–hydraulic analysis.

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.

Validation of the Hybrid Turbulence Model in Detailed Thermal-Hydraulic Analysis Code SPIRAL for Fuel Assembly Using Sodium Experiments Data of 37-Pin Bundles

In the study of safety enhancements on advanced sodium-cooled fast reactors (SFRs) by the Japan Atomic Energy Agency (JAEA), it has been essential to clarify the thermal hydraulics under various operating conditions at high and low flow rate conditions in a fuel assembly (FA) with wire-wrapped fuel pins to assess the structural integrity of the fuel pin that achieves a high-performance core with high burnup ratio and high power density. A finite element thermal-hydraulic analysis code named SPIRAL has been developed by JAEA to analyze the detailed thermal-hydraulic phenomena in the FA of a SFR. In this study, numerical simulations of 37-pin bundle sodium experiments at different Reynolds (Re) number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the developed hybrid k-ε/kθ-εθ turbulence model equipped in SPIRAL to consider the low Re number effect near the wall in the flow and temperature fields. The temperature distributions predicted by SPIRAL were consistent with those measured in the sodium experiments at the Re number conditions. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to the thermal-hydraulic evaluation of sodium-cooled FAs in a wide range of Re numbers was confirmed.

Core flow distribution optimization of a natural circulation reactor using genetic algorithm, simulated annealing and characteristic statistic algorithm

Optimization of core flow distribution is a critical issue in nuclear engineering, particularly for natural circulation reactors with limited driving force. By appropriately arranging the resistance at fuel assembly inlets, core flow distribution could be well matched with core power distribution, thereby reducing the non-uniformity of core outlet temperature distribution. This, in turn, could enhance the performance and safety of the reactor. However, the flow distribution optimization throughout the fuel life cycle is a complicated global optimization problem that entails a vast solution space. In this paper, the flow distribution optimization with minimizing the maximum-to-minimum outlet temperature difference as the primary optimization objective, has been carried out incorporating various meta-heuristic algorithms with the core thermal-hydraulic analysis code COBRA, for a 200 MW natural circulation light water reactor. Minimizing the maximum-to-average outlet temperature difference is also chosen as an optimization objective since a smaller maximum outlet temperature is desirable. After the parametric sensitivity analysis of the genetic algorithm, the simulated annealing, and the characteristic algorithm, a set of parameters of each algorithm for the current flow distribution optimization problem has been determined. The results show that different algorithms yield similar resistance schemes under both optimization objectives, and all exhibit satisfying optimization effects, among which simulated annealing converges to the global optimum with significantly less computational costs.

Evaluation of DNBR with neutronics calculation in LWR systems

The heat flux in a Light Water Reactor (LWR) system is used to estimate the Departure from Nucleate Boiling Ratio (DNBR) of the system which is an important thermal hydraulic parameter for nuclear reactors from heat removal point of view. The DNBR signifies an operational safety limit i.e. the nuclear power plant has to be operated with sufficient margin from the specified DNBR limit for assuring its safety. The DNBR is evaluated using a thermal hydraulic analysis code using inputs from neutronics calculation. The present paper presents the evaluation approach of minimum DNBR (MDNBR) during standard neutronics calculation. The DNBR calculation is performed using a core physics analysis code and burnup variation of MDNBR is studied for the full cycle length. The results of calculation are presented using the equilibrium core of 2700 MWth/900 MWe Indian Pressurized Water Reactor (IPWR). The calculations are performed using VISWAM-TRIHEXFA code system. The few group lattice parametric library for IPWR is generated by lattice analysis code VISWAM. The core follow up calculation for the equilibrium core configuration has been performed using core analysis code TRIHEXFA. A first order thermal hydraulic feedback model has been introduced into the 3D finite difference core simulation tool TRIHEXFA. The critical heat flux calculation, required for estimation of DNBR, has been performed using W-3 Tong and OKB-Gidropress correlations implemented in TRIHEXFA.

Calculation of boron concentration dependent reactivity versus moderator density curve with application on ATWS analysis for Maanshan PWR plant

A method used for determining the moderator density coefficient (MDC) based on the effective multiplication factor (keff) calculated from the SIMULATE-3 code is developed in this study. The calculation is performed at various water density and boron conditions for Maanshan unit 1 cycle 27. The calculated MDC data are fitted as a second-order polynomial function of density, modified with a factor of linear function of boron concentration. The fitted equation is then integrated and normalized to zero at rated density to represent the reactivity as a function of moderator density and boron concentration. A reactivity vs. density table under different boron concentrations can then be generated based on this function. The table generated is used as an input for the ATWS analysis to be performed using the RETRAN-02 thermal–hydraulic analysis code. As an illustration, a table based on a boron concentration of 1500 ppm is generated and used as an input for the analysis of the potential limiting ATWS case, i.e., Loss of Normal Feedwater Flow Event. The ASME emergency level limit of reactor coolant system pressure less than 220.63 bar is used as the acceptance criterion for the current ATWS analysis. It is known that the critical boron concentration at hot full power for each operating fuel cycle varies with the burnup through the cycle. Hence, the calculated ATWS results at different boron concentrations are equivalent to that calculated at different cycle burnup conditions. The cycle burnup point after which the calculated ATWS maximum is within the ASME limit can be determined based on the sensitivity analyses considering different boron concentrations.

A novel method for calculating annular fuel cell resonance effective temperature

Annular fuels can increase the power density of reactors because of their double-sided cooling design, which is of great significance to the miniaturization and long-life operation of pressurized water reactors. In this study, a thermal–hydraulic analysis code named Thermal-Hydraulic Code of Annular Fuel with Single channel (THCAFS) was developed for annular fuels, and a calculation model for the resonance effective temperature of annular fuel cells was established. An annular fuel cell of a four-loop pressurized water reactor (PWR) designed by Westinghouse company was used as the calculation object, and the numerical results of THCAFS were compared with those of existing sub-channel codes such as VIPRE-01, TAFAX, and NACAF. Meanwhile, the Monte Carlo code RMC was used to simulate the radial power distribution and burnup process of the annular fuel, and its thermodynamic behavior was simulated using the self-developed code THCAFS. The radial power distribution, nuclide density variation, and cell temperature field under different burnup levels were also obtained, and the reaction rate equivalence was used instead of the effective resonance integral to investigate the calculation method used for resonance effective temperature of the annular fuel cell. The results show that THCAFS can be preliminarily applied to annular fuel designs and thermal–hydraulic analysis, while the effective temperature calculation method can provide important reference values for studying the resonance effective temperature of annular fuels.

Three-dimensional thermal-hydraulic characteristics analysis of plate-type fuel reactor core based on OpenFOAM

In this paper, a full scale three-dimensional thermal-hydraulic analysis method for plate-type fuel reactor core is proposed through integrating traditional sub-channel analysis concepts into CFD platform. The coupled heat transfer model between fuel plates and narrow rectangular coolant channels is established. The three-dimensional thermal-hydraulic analysis code CorTAF-PT is developed based on open-source CFD platform OpenFOAM. The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. The code simulation results agree well with the experiment and MTR operation data with relative error within 5%. The full-scale MTR simulation under steady state and LOFA is carried out through channel-level resolution. In partial blockage accident, maximum temperature of coolant, fuel pellet and cladding increase by 9.75 K, 60.65 K and 60.52 K respectively. Furthermore, the double temperature peaks of fuel plates and local coolant reflux with −0.121 m s−1 are revealed under partial blockage accident. This work provides reference value for correlational studies on thermal-hydraulic characteristics of plate-type fuel reactor core.

Development and validation of a general two-fluid thermal-hydraulic analysis code for reactors with plate-type fuel assemblies

A general two-fluid thermal-hydraulic analysis code, TTHAC-PFA (Two-fluid Thermal-Hydraulic Analysis Code for reactors using Plate-type Fuel Assemblies), has been developed. The code was developed on basis of the fundamental governing equations. Reasonable mathematic and physical models for parallel rectangular channels and fuel plates have been built up to meet the command of the safety analysis for plate-type fuel reactors. The coolant distribution is conducted on basis of the pressure drop equation theory. Normal subchannel approach is adopted in the code. Fortran language is adopted. The RA-6 experiment is used to evaluate the heat transfer models. The max deviation of the simulations from the measured value is about 5.7 °C. Then, experiments on two-phase flow pressure drop are adopted to validate the code. Approximately 97% of the data points are within ±20% of the deviation range. In addition, CARR (China Advanced Research Reactor) is adopted as the benchmark for normal condition validation. Flow distribution validation and temperature profile validation are conducted. The flow validation presents negligible mean discrepancy and the temperature differences are within the scope of 6 °C. All validations are in good agreement, which indicates the developed code meet the requirements of analyzing thermal-hydraulic characteristics for parallel rectangular channels in reactors. Furthermore, TTHAC-PFA is applied to the investigate channel blockage accident occurred in the CARR. The mass flow is redistributed among channels due to the obstruction. In the obstructed channel and two nearby channels, the coolant temperatures increase a lot. The coolant temperature in the obstructed channel rises by 60 °C. In the rest of the channels, the coolant temperatures decrease with the increase of coolant flow. Besides, the two fuel plates, which form the blocked channel, are asymmetrically cooled due to flow maldistribution. The temperature difference between the cladding on both sides of the fuel plate can reach 40 °C. As a result of partial blockage, the cladding temperature on the side of the blocked channel rises by 63 °C. During the blockage accident, the plate temperature is under the melting temperature. The results indicate the significance of studying the effects of the adjacent channels near the obstructed channel for reactors with plate-type fuel elements.