Code Coupling Research Articles

A tensor-based approach to unify organization and operation of data for irregular spatio-temporal fields

Irregular geographic spatio-temporal-field data have been rapidly accumulating; however, data organizations and operations for different irregular types are often segregated, leading to systematic drawbacks, such as interface expansion difficulty and high coupling codes in GIS implementations. The paper proposes a unified approach to organizing and operating irregular geographic spatio-temporal-field data. The proposed approach has two components, namely ‘concepts and definitions’, and ‘logical model’. The first component introduces the concept of primitive elements, which are formal sets of data points, to serve as the smallest building blocks in the data organization. We define the corresponding primitive elements for three prevalent irregularity types (including sparse, imbalanced, and heterogeneous). The second component utilizes object-oriented programming to support the implementation of various operators. Additionally, we develop the layered architecture to decouple data organization, operation, and visualization to assure low coupling among layers. For demonstrations, we conduct case studies to show the effectiveness of our approach. Additionally, we conduct experiments to new irregularity types and illustrate the flexibility and scalability of our approach. Comparisons with classic tensor methods and spatio-temporal analysis methods show that our approach has more comprehensive supports for different data types.

On the Decoding Performance of Spatially Coupled LDPC Codes With Sub-Block Access

We study spatially coupled LDPC codes that allow access to sub-blocks much smaller than the full code block. Sub-block access is realized by a semi-global decoder that decodes a chosen target sub-block by only accessing the target, plus a prescribed number of helper sub-blocks adjacent in the code chain. This paper develops a theoretical methodology for analyzing the semi-global decoding performance of spatially coupled LDPC codes constructed from protographs. The main result shows that semi-global decoding thresholds can be derived from certain thresholds we define for the single-sub-block graph. These characterizing thresholds are also used for deriving lower bounds on the decoder’s performance over channels with variability across sub-blocks, which are motivated by applications in data storage.

Modeling the process of mixing the coolant in the lower section and pressure mixing chamber of VVER-1200 (V-491)

The most representative accident scenario is determined and a physical and mathematical model is developed for studying the mixing of non-isothermal coolant flows in the structural elements of the V-491 reactor facility (VVER-1200), in which the motion of the medium is described in a three-dimensional non-stationary formulation. Based on the analytical estimates of the list of initiating events, a scenario was chosen with the connection of an idle loop of the main circulation pipeline to three operating ones without a preliminary power reduction. A computational algorithm and a numerical method have been developed for the computational analysis of the selected accident scenario and the justification of the safety of operation of the V-491 reactor facility (VVER-1200). During the numerical simulation, the RANS method was used, which consists in solving the Reynolds-averaged Navier–Stokes equations, the continuity equation and the energy equation. The SST k–ω model of turbulence by Florian Menter is used to close the equations. The verification of the developed physical and mathematical model and calculation procedure was carried out by modeling thermohydraulic processes in models with both a relatively simple geometric design (tee connection) and in a scale model of the reactor vessel (ROCOM experiment), including a lowering section and a pressure mixing chamber. The qualitative agreement of the numerical simulation results with the available data of physical experiments is shown. The results of numerical simulation of the mixing process of non-isothermal coolant flows in the section from the branch pipe of the “cold” thread of the main circulation pipeline to the lower boundary of the fuel of the VVER-1200 core (V-491) are presented. It is shown that the heterogeneity in the temperature distribution at the entrance to the core manifests itself up to 15.5 s of the calculated accident scenario. For calculations the following code coupling were used: Ansys Fluent/Rainbow-TPP.

Development and Testing of TRACE/PARCS ECI Capability for Modelling CANDU Reactors with Reactor Regulating System Response

The use of the USNRC codes TRACE and PARCS has been considered for the coupled safety analysis of CANDU reactors. A key element of CANDU simulations is the interactions between thermal-hydraulic and physic phenomena with the CANDU reactor regulating system (RRS). To date, no or limited development has taken place in TRACE-PARCS in this area. In this work, the system thermal-hydraulic code TRACE_Mac1.0 is natively coupled with the core physic code PARCS_Mac1.0, and RRS control is implemented via the exterior communications interface (ECI) in TRACE. ECI is used for coupling the external codes to TRACE, including additional physical models and control system models. In this work, a Python interface to the TRACE ECI library is developed, along with an RRS model written in Python. This coupling was tested using a CANDU-6 IAEA code coupling benchmark and a 900 MW CANDU model for various transients. For the CANDU-6 benchmark, the transients did not include RRS response, however, the TRACE_Mac1.0/PARCS_Mac1.0 coupling and ECI script functionality was compared to the previous benchmark simulations, which utilized external coupling. For the 900 MW CANDU simulations, all aspects of the ECI module and RRS were included. The results from the CANDU-6 benchmark when using the built-in coupling are comparable to those previously achieved using external coupling between the two codes with coupled simulations taking 2x to 3x less execution time. The 900 MW CANDU simulations successfully demonstrate the RRS functionality for the loss of flow events, and the coupled solutions demonstrate adequate performance for figure-of-eight flow instability modeling.

Calculation of the Natural Frequencies of a Fluid-Filled Pipe System With the Coupled Codes dyvro/rohr2: Implementation of an Alternative Coupling Concept

Abstract The vibration of a pipe system can be caused by internal excitation, e.g., as a result of pump-induced pressure pulsations or external excitation like earthquake. For the investigation of these phenomena, the knowledge about the natural frequencies of the pipe system is essential. In this context, it is important to consider that the natural frequencies of the coupled system of fluid and structure commonly differ from the natural frequencies of the pure fluid system and of the structural pipe model in which the fluid is only taken into account as added mass. To obtain realistic natural frequencies of a pipe system filled with a fluid, the one-dimensional fluid code dyvro was coupled with the code rohr2stoss that deals with dynamic analysis of complex piping and steel structures. Two code coupling concepts were realized, the integral concept, which is also commonly used in coupling commercial pipe system software, and the alternative local coupling concept, with which the longitudinal deformation of the pipe segments can be captured. Both concepts consider a two-way fluid–structure interaction (FSI). For the integral concept, a verification was performed by comparison with an analytical solution for a simple case with a fluid column connected to a spring–mass system. Additionally, both coupling concepts were validated by comparison with experimental data from literature that revealed strong FSI during a water hammer event. The frequency content of the measured pressure fluctuations could be captured by both coupling concepts.

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Abstract The paper investigates the influence of the used thermal-hydraulic approximations on the coupled calculations of gas-cooled fast reactor design (hereby GFR 2400). The NESTLE code is used as coupled simulation tool and solves the multigroup neutron diffusion equation by the finite difference method that is internally coupled with a thermal-hydraulic subchannel code. The in-house developed tempin code and the computational fluid dynamics (CFD) code fluent (from ANSYS code system, Canonsburg, PA) are used to prepare the thermal-hydraulic data for the GFR 2400 calculations. The tempin code solves the steady-state heat balance equation with flowing coolant in triangular lattice cell together with temperature dependent thermal-hydraulic properties of the fuel, cladding, and coolant. Based on the calculated fuel bundle temperature distributions by the tempin code, the thermal-hydraulic material properties (approximations) suitable for the NESTLE coupled code are processed for the GFR 2400 design. The influence of the constant and radial heat generation term within the fuel pin is studied within the paper. The performance of the NESTLE code with thermal-hydraulic approximations processed by both (tempin and fluent) methods is compared with the findings of the GoFastR project. Moreover, both the thermal-hydraulic approximations were compared for one steady-state and one transient state, related to the rapid withdrawal of one control rod assembly from the core. Changes in thermal-hydraulic distributions are described and visualized in the paper.

Inherent Safety Characteristics of Lead Bismuth Eutectic-Cooled Accelerator Driven Subcritical Systems

An accelerator driven subcritical system (ADS) is a new nuclear energy system which could not only produce clean energy but also incinerate nuclear waste. In this paper, inherent safety analysis of an ADS is performed with neutronics and thermal-hydraulics coupled code named ARTAP. Five typical accidents are carried out, including the cases of proton beam interruption, transient overpower, reactivity insertion, loss of flow, and loss of heat sink. The transient simulations are performed in the average channel and the hottest channel of the fuel pin in the ADS core. The simulation results for beam interruption show that the highest temperature of the pellet is in the middle of the fuel element in the average channel, while the peak temperature of the cladding is in the top of the fuel element. After the beam is interrupted for 20s, the maximum temperature drops at the fuel center, the cladding inner surface, and the outlet coolant in the hottest channel are 644.46K, 162.27K, and 136.42K respectively. For transient overpower accidents with the increase of beam intensity, the maximum temperature of the fuel and the cladding are below the safety limit. Concerning the reactivity insertion accident, it is found that the ADS has good inherent safety and its margin of criticality safety is large. The calculation results for loss of flow show that the power drop is small due to low sensitivity of the subcritical core to negative reactivity feedback, and the maximum temperature of the cladding reaches 1726K, which means the fuel element would rupture. However, the power and the temperatures of fuel, cladding, and coolant could decrease quickly to the safety level after the proton accelerator is cut off under a loss of flow accident. The results also show that the peak temperature of the cladding is lower than the safety limit under a loss of heat sink accident. The present simulation results reveal that the ADS has a remarkable advantage against severe accidents. It also implies that its inherent safety characteristics could ensure reactor shutdown by cutting off the proton beam after accidents occur.

Associated Sectors of Magnetic Recording Systems Using Spatially Coupled LDPC Codes

In traditional magnetic recording systems, non-associated sectors are mainly adopted, whereby two consecutive sectors are decoded independently by the low-density parity-check (LDPC) codes. In this paper, we propose a magnetic recording system with associated sectors, constructed using spatially coupled low-density parity-check (SC-LDPC) codes. If the SC-LDPC decoder cannot correct the erroneous bits in the current sector, it can request information stored in previous sectors to improve decoding performance. Moreover, we modify protograph-based extrinsic information transfer (P-EXIT) charts to examine the theoretical performance of SC-LDPC codes applied to both non-associated and associated sectors. Our theoretical results show that the associated sectors achieve significant performance gains compared to the traditional non-associated sectors.

Coupled neutronics and thermal-hydraulics transient simulation of a gas-cooled reactor in the aircraft nuclear propulsion system

Aircraft nuclear propulsion (ANP) has many historical explorations, but also features with high power density and high operating temperature. These challenges lead to numerical investigations for the transient behaviors of gas-cooled reactors to be used for ANP system. This prismatic-block type reactor is featured by a large number of internal discrete cooling holes with high aspect ratio. When using conjugate heat transfer (CHT) model in CFD analyses for the entire solid core of such type of reactors, the refined mesh resolution near the interface between the fuel matrix and the flow channels leads to high computational costs. Therefore, we proposed a new porous heat transfer (PHT) model with Newton’s law of cooling to predict three-dimensional distributions of temperature in the solid and fluid domain of the reactor core. The verification of the PHT model is conducted in the presented study firstly, and shows that the accuracy of this model is deeply influenced by the prediction of convective heat transfer coefficients. Sequentially, a neutronics (N) and thermal-hydraulics (TH) transient coupling code is developed in the open-source platform OpenFOAM. The point kinetic model is employed to evaluate the dependency of heat power on the time. Doppler broadening effect is considered in the coupling process by delivering fuel average temperatures. Then coupled N/TH simulations are carried out to analyze the gas-cooled reactor during the start-up process and the blockage accident. The obtained results show that the wall temperature is always lower than the design limit even in a rapid start-up process, indicating that the reactor is safe at extreme operations. Moreover, simulation of the blockage accident shows that if several flow channels are totally occupied by debris, the fuel temperature increment is less than 300 K, and the position of the peak fuel temperature transforms from the core center to the zone near the blockage.

Concatenated Spatially Coupled LDPC Codes With Sliding Window Decoding for Joint Source-Channel Coding

In this paper, a method for joint source-channel coding (JSCC) based on concatenated spatially coupled low-density parity-check (SC-LDPC) codes is investigated. A construction consisting of two SC-LDPC codes is proposed: one for source coding and the other for channel coding, with a joint belief propagation-based decoder. Also, a novel windowed decoding (WD) scheme is presented with significantly reduced latency and complexity requirements. The asymptotic behavior for various graph node degrees is analyzed using a protograph-based Extrinsic Information Transfer (EXIT) chart analysis for both LDPC block codes with block decoding and for SC-LDPC codes with the WD scheme, showing robust performance for concatenated SC-LDPC codes. Simulation results show a notable performance improvement compared to existing state-of-the-art JSCC schemes based on LDPC codes with comparable latency and complexity constraints.

The strip method for shape derivatives

AbstractA major challenge in shape optimization is the coupling of finite element method (FEM) codes in a way that facilitates efficient computation of shape derivatives. This is particularly difficult with multiphysics problems involving legacy codes, where the costs of implementing and maintaining shape derivative capabilities are prohibitive. The volume and boundary methods are two approaches to computing shape derivatives. Each has a major drawback: the boundary method is less accurate, while the volume method is more invasive to the FEM code. We introduce the strip method, which computes shape derivatives on a strip adjacent to the boundary. The strip method makes code coupling simple. Like the boundary method, it queries the state and adjoint solutions at quadrature nodes, but requires no knowledge of the FEM code implementations. At the same time, it exhibits the higher accuracy of the volume method. As an added benefit, its computational complexity is comparable to that of the boundary method, that is, it is faster than the volume method. We illustrate the benefits of the strip method with numerical examples.

Development of Coupled PROTEUS and SAM Code System for Multiphysics Analysis of Advanced Reactors

Development of Coupled PROTEUS and SAM Code System for Multiphysics Analysis of Advanced Reactors

Stark–Zeeman line-shape model for multi-electron radiators in hot dense plasmas subjected to large magnetic fields

We present a Stark–Zeeman spectral line-shape model and the associated numerical code, PPPB, designed to provide fast and accurate line shapes for arbitrary atomic systems for a large range of plasma conditions. PPPB is based on the coupling of the PPP code—a Stark-broadened spectral line-shape code developed for multi-electron ion spectroscopy in hot dense plasmas—and the MASCB code developed recently to generate B-field-dependent atomic physics. The latter provides energy levels, statistical weights, and reduced matrix elements of multi-electron radiators by diagonalizing the atomic Hamiltonian that includes the well know B-dependent term. These are then used as inputs to PPP working in the standard line-broadening approach, i.e., using the quasi-static ion and impact electron approximations. The effects of ion dynamics are introduced by means of the frequency fluctuation model, and the physical model of electron broadening is based on the semi-classical impact approximation including the effects of a strong collision term, interference, and cyclotron motion. Finally, to account for polarization effects, the output profiles are calculated for a given angle of observation with respect to the direction of the magnetic field. The potential of this model is presented through Stark–Zeeman spectral line-shape calculations performed for various experimental conditions.

Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel

To accurately assess the in-pile performance of the accident tolerant fuel (ATF) over the whole life for the light water reactors (LWR), a full coupling of neutronics/thermal-hydraulics/fuel-performance is needed. In this paper, a multiphysics coupling code system named NECP-CLAMPERL is developed based on the high-fidelity neutronics code NECP-X, the sub-channel thermal-hydraulics code CTF, the finite-element fuel performance code NECP-CALF and the multiphysics object-oriented simulation environment (MOOSE). The code system is then applied to evaluate the performance of U3Si2 fuel, one of the most promising ATFs, by simulating a typical fuel assembly in pressurized water-cooled reactors (PWR). The quantities of interest including the reactivity, power distribution, fuel temperature, clad stress, fission gas release, plenum pressure, fuel radial displacement, gap width and PCMI are presented and analyzed in comparison with the traditional UO2 fuel. Future works and other potential applications are also identified.

RayActive: An all-in-one R2S neutron induced activation analysis tool based on new intrinsic resolution methods.

Neutron production in nuclear facilities can induce a SDDR (Shutdown Dose Rate) even after months. Different methods exist to simulate the SDDR of neutron activated materials. The most used method is the mesh-based R2S (Rigorous-Two-Steps). This method couples a Monte-Carlo code and an isotopic inventory code to compute the SDDR. The Monte-Carlo code computes the neutron flux on a user-defined mesh to take into account the neutron flux gradients in the cells. The aim of this article is to present RayActive, a prototype tool elaborated for computing the SDDR. RayActive is based on a modified R2S method with an almost exact identification of materials enclosed in each voxel of a superimposed mesh. The CAD based geometry definition for the model makes possible this identification. Moreover, RayActive is designed as an ”all-in-one” package without any other code coupling. RayActive is designed to be as easy to use as possible and each part of the computation has been optimized to minimize the computational time. Verification and validation of RayActive has been done on shutdown dose rate benchmarks made with FLUKA code and classical R2S process coupling MCNP and FISPACT codes. Good agreement was found for these benchmarks even if some differences occur for the validation benchmark. These discrepancies show the impact on the results of the modified R2S method employed in RayActive.

Status and Progress of the Multi-Physics Coupling and Multi-Scale Coupling Research for Numerical Reactors

This paper explains the basic concept of numerical reactors, and studies in detail the international research and development (R&D) projects for numerical reactors, such as the Consortium for Advanced Simulation of Light Water Reactors (CASL), European Nuclear Reactor Simulation (NURESIM) platform, and Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. It also presents a further study on and a summary of the current research status of multi-physics coupling and multi-scale coupling technology in China and other countries. In combination with the research status, this paper indicates that the focus of the numerical reactor technology development rests on the multi-physics coupling mechanism under the combined action of material corrosion, flow heat transfer and neutronics, and the development of high-fidelity coupling code based on unified grid solution.

Comparative Safety Analysis of Accelerator Driven Subcritical Systems and Critical Nuclear Energy Systems

The accelerator driven subcritical system (ADS) has been chosen as one of the best candidates for Generation IV nuclear energy systems which could not only produce clean energy but also incinerate nuclear waste. The transient characteristics and operation principles of ADS are significantly different from those of the critical nuclear energy system (CNES). In this work, the safety characteristics of ADS are analyzed and compared with CNES by a developed neutronics and thermal-hydraulics coupled code named ARTAP. Three typical accidents are carried out in both ADS and CNES, including reactivity insertion, loss of flow, and loss of heat sink. The comparison results show that the power and the temperatures of fuel, cladding, and coolant of the CNES reactor are much higher than those of the ADS reactor during the reactivity insertion accident, which means ADS has a better safety advantage than CNES. However, due to the subcriticality of the ADS core and its low sensitivity to negative reactivity feedback, the simulation results indicate that the inherent safety characteristics of CNES are better than those of ADS under loss of flow accident, and the protection system of ADS would be quickly activated to achieve an emergency shutdown after the accident occurs. For the loss of heat sink, it is found that the peak temperatures of the cladding in the ADS and CNES reactors are lower than the safety limit, which imply these two reactors have good safety performance against loss of heat sink accidents.

High-fidelity multi-physics coupling study on advanced heat pipe reactor

The advanced heat pipe cooled reactor is a potential candidate to generate nuclear power for space exploration, and exhibits strong neutron leakage and complicated multi-physics coupled effects. Traditional numerical methods based on single-field simulations do not adequately describe these interactive physical phenomena. It is thus necessary to apply high-fidelity multi-physics coupling simulations for the analysis and design of heat pipe reactors. In this work, a three-dimensional high-fidelity neutronics-thermo-elasticity multi-physics coupling code is developed for the heat pipe reactor, Kilowatt Reactor Using Stirling TechnologY (KRUSTY), based on the Monte Carlo method and the finite element method. The code combines existing open-source codes (OpenMC, Nektar++, SfePy) and implements the functional expansion tally method to perform data mapping between the Monte Carlo and the finite element method solver.An on-the-fly convergence criterion dedicated to the functional expansion tally method is developed based on statistical uncertainties and the L2 norm. This new criterion is shown to be unconditionally stable with various statistical uncertainties. Using the coupling code, high-fidelity coupling simulations are performed under different steady-state conditions, which provide insights to the physical phenomena of the reactor. The coupling results present same trends as the previous KRUSTY simulation, and illustrate that the feedback from thermal expansion is critical for capturing the negative reactivity feedback. The heat pipe analysis shows that a sufficient heat-removing margin for heat pipe failure accidents is guaranteed. The depletion coupling result shows that the burn-up effects are negligible for the reactor.

CFD/RELAP5 coupling analysis of the ISP No. 43 boron dilution experiment

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

Efficient Coupling of Hydrodynamic and Energy-Deposition Codes for Hydrodynamic-Tunnelling Studies on High-Energy Particle Accelerators

The machine-protection evaluation of high-energy accelerators comprises the study of beyond-design failures, including the direct beam impact onto machine elements. In case of a direct impact, the nominal beam of the Large Hadron Collider (LHC) would penetrate more than 30 meters into a solid copper target. The penetration depth due to the time structure of the particle beam is, thus, significantly longer than predicted from purely static energydeposition simulations with 7 TeV protons. This effect, known as hydrodynamic tunnelling, is caused by the beam-induced density depletion of the material at the target axis, which allows subsequent bunches to penetrate deeper into the target. Its proper simulation requires, therefore, to sequentially couple an energy-deposition code and a hydrodynamic code for the different target densities. This paper describes a method to efficiently couple the simulations codes Autodyn and FLUKA based on automatic density assignment and input file generation, and presents the results achieved for a sample case.