Code Coupling Research Articles

Two-Phase Globally Coupled Low-Density Parity Check Decoding Aided with Early Termination and Forced Convergence.

To enhance the decoding efficiency of Globally Coupled (GC) LDPC codes, we incorporated Early Termination (ET) and Forced Convergence (FC) into the local/global two-phase decoding algorithm to expedite the decoding process. The two-phase decoding scheme integrates the ET technique to halt unnecessary iterations in the local decoding phase while employing the FC technique to accelerate convergence in the global phase decoding. The application of ET technology in the local decoding of GC-LDPC codes will not cause performance loss as in traditional block codes and will cause considerable complexity gains. For a longer code length and larger convergence differences between nodes' global codes, the FC technique operates more efficiently in global code than local code. Two variants are proposed for the ET scheme in the local decoding, namely ET-1 and ET-2. The initial variant, ET-1, predicts whether local decoding can be successful according to data characteristics and stop the local decoding iteration that is not expected to be successful in time. In the case of ET-2, the saved local iterations are transformed to global decoding equally. The results show that ET-1 saves considerable decoding time complexity and ET-2 improves the performance of the GC-LDPC code with the same decoding time complexity. The combined approach of ET-1 with FC reduces the decoding time complexity up to 42% at a low Signal Noise Rate region while maintaining its performance; ET-2-FC two-phase decoding saves approximately 25% decoding time complexity while improving the BER by about 0.18 dB and FER by about 0.23 dB.

Verification of direct coupling code system using FRENDY version 2 and GENESIS for light water reactor lattices

ABSTRACT This study newly established a direct coupling code system consisting of the nuclear data processing code FRENDY version 2, and the three-dimensional heterogeneous transport code GENESIS (FRENDY-V2/GENESIS) for easy implementation of the random-sampling-based uncertainty quantification considering the implicit effect due to nuclear cross-section (XS) perturbations. The multi-group macroscopic XSs prepared for GENESIS were generated by FRENDY version 2, where the Dancoff factor was calculated by the neutron current method. Then the background XSs were evaluated based on the Carlvik two-term rational approximation. The infinite multiplication factor (k-infinity) and the fission reaction rate distribution in UO2 and MOX lattice geometries were compared with MVP3 to verify the calculation accuracy of FRENDY-V2/GENESIS. The sensitivity analyses on the discretization conditions such as the ray tracing of the method of characteristics were also carried out. Through several comparisons between FRENDY-V2/GENESIS and MVP3, FRENDY-V2/GENESIS with the SHEM 361-group structure calculates the k-infinity within approximately 50 pcm and the fission reaction rate distribution within approximately 0.1% by the root mean square, respectively. Consequently, the applicability of FRENDY-V2/GENESIS was verified, and FRENDY-V2/GENESIS can be used to discuss the implicit effect due to multi-group XS perturbations.

Preliminary study of transuranic transmutation in a small modular chloride salt fast reactor

The management of radioactive waste poses a significant challenge to the sustainable development of nuclear energy. Efficient transmutation of nuclear wastes is crucial to minimize their accumulation. A small modular chloride salt fast reactor (sm-MCFR) capable of transmuting transuranic elements (TRU) is proposed in this paper, combining the advantages of the small modular reactor (SMR) and the molten salt reactor (MSR). The sm-MCFR is characterized by a high fuel loading and a compact core structure that can be quickly deployed around large commercial reactors to achieve TRU transmutation. To evaluate the TRU burnup capability of the sm-MCFR, several fuel salts and reprocessing modes were analyzed using the internally developed TRITON MODEC Coupled Burnup Code (TMCBurnup) tool. NaCl-MgCl3 with 98 % enrichment in 37Cl is chosen as carrier salt for the sm-MCFR, which can achieve 76.7 % TRU transmutation rate in average and 355 kg·GW−1·a−1 TRU transmutation quality at a continuous reprocessing rate of 10 L/d for 50 operation years. The optimized sm-MCFR reduced the radioactive toxicity of TRU by 84 %, thereby simplifying waste reprocessing. In addition, the sm-MCFR has a negative temperature feedback coefficient of −7.195 pcm/K, favoring safe reactor operation.

Development and validation of ARSAC-CORTH coupling code based on a generic coupling architecture

Development and validation of ARSAC-CORTH coupling code based on a generic coupling architecture

Coupling of the best-estimate system code and containment analysis code and its application to TMLB’ accident

With the development of advanced pressurized water reactor technology, the thermal-hydraulic coupling effect between the containment and the primary system becomes increasingly tight. In order to meet the demand for integrated safety analysis between the containment and the primary system, this paper investigates a direct coupling method between the best-estimate system code Advanced Reactor Safety Analysis Code and the containment analysis program ATHROC (Analysis of Thermal Hydraulic Response Of Containment). The feasibility of this direct coupling method and the applicability of the coupled program for overall safety analysis are demonstrated using Marviken two-phase flow release experiments. The ATHROC/ARSAC coupled program is employed to analyze the impact of the pressure relief function of the CPR1000 nuclear power plant pressurizer on the behavior of the primary system and containment during the TMLB’ accident. The calculation results indicate that these measures can reduce the pressure of the primary system to the level acceptable by the low-pressure injection system, but at the same time, they cause the pressure in the containment to rise to nearly 0.4 MPa. Therefore, to ensure the structural integrity of the containment, it is necessary for the non-passive hydrogen recombiner to effectively reduce the hydrogen concentration, thereby avoiding additional pressure increase in the containment due to hydrogen deflagration, which could lead to overpressure failure. The findings of this study are of significant reference value for improving the safety performance of thermal-hydraulic systems in operational Gen-II and advanced Gen-III pressurized water reactor nuclear power plants.

Effect of Fluid-Solid-Thermal Coupling on the Performance of S-Duct in Hypersonic Inlet

This article studies the effects of multi-field coupling on the S-duct by a self-developed fluid-solid-thermal bi-directional coupling code. The flow field inside the S-duct is complex with various vortices as the airflow turns. The results indicate that the total pressure recovery coefficient and the total pressure distortion index (DI) change rapidly at the beginning and reach a steady state at about 100s. The σ decreases by 12%, and the DI increases by 2% compared to the starting point. Structural deformation, size of separation zone, and wall temperature changes over time, and stabilizes around 100 s. The deformation mainly occurs at the outlet, with a maximum displacement of 3.65 mm for up-warping and 4.60 mm for expansion. The values of the total pressure recovery coefficient with S-ducts of bias distance with 20, 40, and 60 mm drop by 2.5%, 8.6%, and 12.1%, and the values of the total pressure distortion index increases with S-ducts of bias distance with 20, 40, and 60 mm increase by 10.8%, 15.1%, and 1.7%. The S-duct with a larger aspect ratio has larger separation zones and a smaller total pressure recovery coefficient.

Assessing the impact of DIONISIO-SubChanFlow code coupling in nuclear fuel performance simulations

Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.

Four-Cycle Free Spatially Coupled LDPC Codes With an Explicit Construction

Four-Cycle Free Spatially Coupled LDPC Codes With an Explicit Construction

A multi-scale and multi-physical coupling method for the transient characteristics of space nuclear reactor

In the future of deep space exploration, space nuclear reactors (SNRs) are essential because they can provide a robust, sustainable energy supply independent of the environment. SNRs have complex multi-physics coupling effects, and their systems involve many modules. The traditional method uses the lumped parameter model, or single-filed, which cannot accurately describe the transient characteristics and has low spatial resolution and poor data representation. Thus, developing a multi-dimensional, multi-physics coupled method to analyze and design SNRs is necessary. This work proposes a multi-dimensional, multi-physics coupled method for the SNRs. A transient coupling code for the space thermionic reactor TOPAZ-II based on OpenFOAM was developed. The reactor neutron physics model, multi-region heat transfer model, thermoelectric conversion model, electromagnetic pump model, and heat pipe radiator model are included. Different scale models are imported by the code, including a three-dimensional reactor model and a two-dimensional heat pipe. Furthermore, the experimental data from the V-71 test unit were used to verify the developed code. The shutdown and startup of the improved TOPAZ-II system are simulated using the developed code. The results show that the calculated results agree well with the experimental data. During the startup and shutdown processes, the material temperatures remained within acceptable limits, so it can be considered that the developed 2D/3D system analysis for SNRs has successfully achieved a high-dimensional simulation of transient characteristics, accurately capturing the spatial distribution of SNRs at different scales.

Two sides of the same coin: A study on developers' perception of defects

SummarySoftware defect prediction is a subject of study involving the interplay of software engineering and machine learning. The current literature proposed numerous machine learning models to predict software defects from software data, such as commits and code metrics. Further, the most recent literature employs explainability techniques to understand why machine learning models made such predictions (i.e., predicting the likelihood of a defect). As a result, developers are expected to reason on the software features that may relate to defects in the source code. However, little is known about the developers' perception of these machine learning models and their explanations. To explore this issue, we focus on a survey with experienced developers to understand how they evaluate each quality attribute for the defect prediction. We chose the developers based on their contributions at GitHub, where they contributed to at least 10 repositories in the past 2 years. The results show that developers tend to evaluate code complexity as the most important quality attribute to avoid defects compared with the other target attributes such as source code size, coupling, and documentation. At the end, a thematic analysis reveals that developers evaluate testing the code as a relevant aspect not covered by the static software features. We conclude that, qualitatively, there exists a misalignment between developers' perceptions and the outputs of machine learning models. For instance, while machine learning models assign high importance to documentation, developers often overlook documentation and prioritize assessing the complexity of the code instead.

Verification of the Coupled Code PARCS/TWOPORFLOW with Rod Ejection Accident Calculations for Small Modular Reactors

Research on small modular reactors (SMRs) is gaining importance since they are key for addressing energy challenges in various sectors. These types of reactors integrate novel technologies that rely heavily on passive safety systems. Among the most developed light water reactors, SMR designs are SMART and NuScale. This work analyzes the academic boron-free Karlsruhe Small Modular Reactor (KSMR), which may fit in SMART, and a core design resembling NuScale. The research aims to explore the potential of new multiphysics tools under development to predict safety parameters of SMRs during normal operation and transients. For this purpose, the Purdue Advanced Reactor Core Simulator (PARCS) and TWOPORFLOW (TPF) codes have been coupled using the Interface for Code Coupling (ICoCo), orchestrating code execution through a C++ Supervisor program. In this work, a nodewise rod ejection accident (REA) has been analyzed for two different scenarios. The first is a hot-zero-power scenario for the KSMR core, and the second is initiated at 75% of nominal power for the NuScale core. Verification of results has been done though code-to-code comparison. Comparison of the PARCS/TPF results obtained for both KSMR and NuScale with reference cases shows acceptable differences. Key safety parameters predicted by the codes for the REA analysis of both cores have also been evaluated against the latest U.S. Nuclear Regulatory Commission regulation for reactivity initiated accidents showing that all parameters fulfill core coolability acceptance criteria and fuel rod cladding failure thresholds. The presented work highlights the transition from the coupling of PARCS with a mixture model code, such as SCF, to a coupling with a two-phase model code, such as TPF. These findings contribute to a better understanding of SMR phenomenology during accidental sequences and demonstrate the capabilities of the coupled codes.

Optimizing Coupled Fluid-Structure Simulations for Nuclear-Relevant Geometries

Abstract The numerical simulation of fluid-structure interactions (FSI) has gained interest to study flow-induced vibrations. Nevertheless, the high computational resources required by such simulations can represent a significant limitation for their application to industrial configurations. Therefore, simplified modeling approaches, when physically applicable, can represent an interesting compromise. This can be the case of slender structures (tubes, rods) often encountered in nuclear power plants. In this paper, an Euler–Bernoulli beam finite element model is implemented inside the computational fluid dynamics (CFD) code code_Saturne. With the goal of finding CFD methods less expensive than large eddy simulations (LES), unsteady Reynolds Navier–Stokes (URANS) and hybrid URANS/LES approaches are considered. The resulting fluid-structure model is able to calculate the vibration response of cantilever beams under a fluid flow, avoiding the necessity of CFD-finite element method (FEM) code coupling. The first part of the paper describes the model and its implementation: it allows to perform 2-way explicit fluid-structure coupling, using the Arbitrary Lagrangian-Eulerian approach to account for the structure deformations. Validation test cases are presented in the second part: first, the model is validated in terms of frequency, added mass, and damping for a cylinder vibrating in static air and water; then, the model is validated toward the vortex-induced resonance and lock-in mechanisms for a cylinder subjected to water cross-flow. The model is then applied to a real experimental configuration of two in-line cylinders in water cross-flow: the calculated vibrations are found to be in good agreement with the experimental measurements.

Fast ion confinement in the presence of core magnetic islands in Wendelstein 7-X

The effect of magnetic islands in the core region of Wendelstein 7-X (W7-X) on fast ion confinement is explored through simulations with the BEAMS3D code. A magnetic configuration where the n/m=5/5 island chain is shifted to r/a∼0.7 allows the exploration of core island physics in W7-X. The control coil system on W7-X allows the tuning of the island size either increasing the island width or decreasing it. A coupling of the BEAMS3D code to the FIELDLINES code provides a versatile mechanism for incorporating magnetic islands and stochastic regions into the BEAMS3D code. Collisionless simulations suggest that the presence of core islands degrade the confinement of passing particles in the region of the island chain. Full neutral beam simulations of W7-X show a similar behavior with confinement decreasing as the island width is increased. Comparisons between a vacuum magnetic field and low beta HINT2 simulation are made showing similar fast ion behavior. Measurements of lost fast ions in W7-X confirm this trend with the control coil suppressed island configuration showing lower losses than that with no control coils applied. Simulations of fast ion wall loads are performed suggesting no drastic change in loss pattern and a slight reduction in losses with minimized islands.

Spatially Coupled Codes via Bidirectional Block Markov Superposition Transmission

Spatially Coupled Codes via Bidirectional Block Markov Superposition Transmission

Physical design and recent experimental results of the new ICRF antenna on EAST

Two new ICRF antennas operating in the ion cyclotron radio frequency (ICRF) range have been developed for EAST to overcome the low coupling problem of the original antennas. The original ICRF antennas were limited in their power capacity due to insufficient coupling. The new antenna design takes into account both wave coupling and absorption processes through comprehensive wave coupling and absorption codes, with the dominant parallel wave number of 7.5 m−1 at dipole phasing. Through the use of these new ICRF antennas, we are able to achieve 3.8 MW output power and 360 s operation, respectively. The initial experimental results demonstrate the reliability of the antenna design method.

Hemodynamics and wall shear metrics in a pulmonary autograft: Comparing a fluid-structure interaction and computational fluid dynamics approach

ObjectiveIn young patients, aortic valve disease is often treated by placement of a pulmonary autograft (PA) which adapts to its new environment through growth and remodeling. To better understand the hemodynamic forces acting on the highly distensible PA in the acute phase after surgery, we developed a fluid-structure interaction (FSI) framework and comprehensively compared hemodynamics and wall shear-stress (WSS) metrics with a computational fluid dynamic (CFD) simulation. MethodsThe FSI framework couples a prestressed non-linear hyperelastic arterial tissue model with a fluid model using the in-house coupling code CoCoNuT. Geometry, material parameters and boundary conditions are based on in-vivo measurements. Hemodynamics, time-averaged WSS (TAWSS), oscillatory shear index (OSI) and topological shear variation index (TSVI) are evaluated qualitatively and quantitatively for 3 different sheeps. ResultsDespite systolic-to-diastolic volumetric changes of the PA in the order of 20 %, the point-by-point correlation of TAWSS and OSI obtained through CFD and FSI remains high (r > 0.9, p < 0.01) for TAWSS and (r > 0.8, p < 0.01) for OSI). Instantaneous WSS divergence patterns qualitatively preserve similarities, but large deformations of the PA leads to a decrease of the correlation between FSI and CFD resolved TSVI (r < 0.7, p < 0.01). Moderate co-localization between FSI and CFD is observed for low thresholds of TAWSS and high thresholds of OSI and TSVI. ConclusionFSI might be warranted if we were to use the TSVI as a mechano-biological driver for growth and remodeling of PA due to varying intra-vascular flow structures and near wall hemodynamics because of the large expansion of the PA.

Coupling between LOTUS and CTF with DYN3D within a multiscale and multiphysics software development

Coupling of nuclear codes can be performed at several scale levels and is necessary to improve their reliability and sustainability. Mainly, the coupling of nuclear codes that use nodal expansion neutronics with channel thermal hydraulics has been carried out at the fuel assembly level while, only recently, the coupling of nuclear codes that use advanced neutronics, thermal hydraulics, and thermo-mechanics has been carried out at either the fuel pin or materials level. Meanwhile, in the UK, a multiscale and multi-physics software development between NURESIM and CASL is being developed, which includes a coupling software environment that enables the coupling of nuclear codes at several scale levels. Full coupled reactor physics at either the fuel pin or materials level can be obtained by coupling the transport code LOTUS, the subchannel code CTF, and the nodal code DYN3D. In this journal article, a multi ways coupling between LOTUS and CTF at either the fuel pin or materials levels with DYN3D at the fuel assembly level is compared to a multi ways coupling between DYN3D and CTF at the fuel pin level with DYN3D at the fuel assembly level and a multi ways coupling between Open MC and CTF at the materials level with DYN3D at the fuel assembly level. These comparisons have been carried out to present the coupled reactor physics verifications at either the fuel pin or materials levels. These show that the multi ways coupling between LOTUS and CTF at either the fuel pin or materials levels with DYN3D at the fuel assembly level outperforms the multi ways coupling between DYN3D and CTF at the fuel pin level with DYN3D at the fuel assembly level due to the application in the former of full neutron transport. Also, these show that the multi ways coupling between LOTUS and CTF at the materials level with DYN3D at the fuel assembly level agrees with the multi ways coupling between Open MC and CTF at the materials level with DYN3D at the fuel assembly level due to the application in both of full neutron transport.

Coupled 3D thermal-hydraulic code development for performance assessment of spent nuclear fuel disposal system

As a solution to the problem of spent nuclear fuels (SNFs), the disposal of SNF has gained attention from nations using nuclear energy because of hazards posed to the ecosystem. Among many proposed solutions, the most promising method is to dispose of SNF in a deep geological repository (DGR) which utilizes the multi-barrier concept developed by Finland and Sweden. Here, a new fully-coupled Thermal-Hydraulic (TH) code HADES (High-level rAdionuclide Disposal Evaluation Simulator) is developed using the MOOSE framework. This new code suggests basic numerical tools, such as a non-linear solver and finite element discretization, to assess the safety performance of disposal systems. The new TH code considered various TH behavior using Richards' flow approach, assuming gas pressure is constant. The HADES showed promising results when it was compared to various TH codes validated from DECOVAELX-THMC projects. When the single-canister model was utilized to estimate the TH behavior of the Korean Reference disposal System, although it showed significant saturation reduction due to the evaporation of water, the temperature was maintained under the thermal criteria limit, which is 100 °C. In addition, the new code estimated temperature and degree of saturation of the multi-canisters model, considering two or three canisters, it showed a slightly lower temperature, 5 °C, than the single-canister model. From these results, the following are concluded: (1) the new TH code contribute to an additional integrity by estimating TH behavior of KRS; (2) however, due to limitations in single-canister simulation, it is recommended to use multi-canisters simulation to estimate TH behavior accurately. Therefore, this model is anticipated not only to help licensing applications and estimation of various multi-physics phenomena and multi-canister at the disposal site.

Multi-physics code coupling methodology for the simulation of accidental transients in sodium cooled fast reactors

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures).In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years.This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA’s Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided.The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident.This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation.

Computational study of the Nitrogen-16 source term in the ITER vacuum vessel cooling circuit through the coupling of system-level analysis code and CFD

In ITER, the evaluation of the activated water radiation source and its impact on the radiological levels is necessary to demonstrate compliance with the safety requirements. The use of simplified or conservative approaches often results in the application of expensive constraints on the installation that impact its economics, operations, and construction schedule. In this work, we propose a novel methodology to calculate the activated water source term with a higher degree of realism. The methodology is based on the coupling of a system-level code with a Computational Fluid Dynamics (CFD) code in an explicit, one-way approach. We apply this methodology to the evaluation of the 16N radioisotope within the ITER Vacuum Vessel Primary Heat Transfer System (VV-PHTS) cooling circuit in a steady-state and transient scenarios. We chose this system since previous analyses of the VV-PHTS were done with simple, ad-hoc calculations that yielded results that differed by up to a factor of five, underscoring a higher level of uncertainty. As a result, we generate a computational model of the source term that can be used to evaluate the radiological condition surrounding the cooling systems during the operations.