scholarly journals High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

2014 ◽  
Vol 372 (2021) ◽  
pp. 20130381 ◽  
Author(s):  
Vijay S. Mahadevan ◽  
Elia Merzari ◽  
Timothy Tautges ◽  
Rajeev Jain ◽  
Aleksandr Obabko ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Neutron Transport ◽  
High Fidelity ◽  
Design Problems ◽  
Flexible Coupling ◽  
Physics Simulation ◽  
Software Interfaces ◽  
Coupling Strategy ◽  
Reactor Models ◽  
Computational Resources

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

scholarly journals Innovations in Multi-Physics Methods Development, Validation, and Uncertainty Quantification

2021 ◽  
Vol 2 (1) ◽  
pp. 44-56
Author(s):  
Maria Avramova ◽  
Agustin Abarca ◽  
Jason Hou ◽  
Kostadin Ivanov
Keyword(s):  
Uncertainty Quantification ◽  
Spatial Scale ◽  
Data Science ◽  
Nuclear Reactor ◽  
High Fidelity ◽  
Energy Agency ◽  
Multi Scale ◽  
Physics Simulation ◽  
Safety Parameters ◽  
Impact Problems

This paper provides a review of current and upcoming innovations in development, validation, and uncertainty quantification of nuclear reactor multi-physics simulation methods. Multi-physics modelling and simulations (M&S) provide more accurate and realistic predictions of the nuclear reactors behavior including local safety parameters. Multi-physics M&S tools can be subdivided in two groups: traditional multi-physics M&S on assembly/channel spatial scale (currently used in industry and regulation), and novel high-fidelity multi-physics M&S on pin (sub-pin)/sub-channel spatial scale. The current trends in reactor design and safety analysis are towards further development, verification, and validation of multi-physics multi-scale M&S combined with uncertainty quantification and propagation. Approaches currently applied for validation of the traditional multi-physics M&S are summarized and illustrated using established Nuclear Energy Agency/Organization for Economic Cooperation and Development (NEA/OECD) multi-physics benchmarks. Novel high-fidelity multi-physics M&S allow for insights crucial to resolve industry challenge and high impact problems previously impossible with the traditional tools. Challenges in validation of novel multi-physics M&S are discussed along with the needs for developing validation benchmarks based on experimental data. Due to their complexity, the novel multi-physics codes are still computationally expensive for routine applications. This fact motivates the use of high-fidelity novel models and codes to inform the low-fidelity traditional models and codes, leading to improved traditional multi-physics M&S. The uncertainty quantification and propagation across different scales (multi-scale) and multi-physics phenomena are demonstrated using the OECD/NEA Light Water Reactor Uncertainty Analysis in Modelling benchmark framework. Finally, the increasing role of data science and analytics techniques in development and validation of multi-physics M&S is summarized.

scholarly journals SERPENT/SUBCHANFLOW COUPLED CALCULATIONS FOR A VVER CORE AT HOT FULL POWER

2021 ◽  
Vol 247 ◽  
pp. 04006
Author(s):  
Diego Ferraro ◽  
Manuel García ◽  
Uwe Imke ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
...  
Keyword(s):  
High Performance ◽  
Neutron Transport ◽  
Coupling Scheme ◽  
High Fidelity ◽  
Reactor Physics ◽  
Coupled Calculation ◽  
Final Goal ◽  
Wide Range ◽  
Computational Resources ◽  
Full Core

An increasing interest on the development of highly accurate methodologies in reactor physics is nowadays observed, mainly stimulated by the availability of vast computational resources. As a result, an on-going development of a wide range of coupled calculation tools is observed within diverse projects worldwide. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes. These tools are aimed to be suitable for high-fidelity calculations both for PWR and VVER reactors, with the final goal of performing pin-by-pin coupled calculations at full core scope including burnup. Several intermediate steps are to be analyzed in-depth before jumping into this final goal in order to provide insights and to identify resources requirements. As part of this process, this work presents the results for a pin-by-pin coupling calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel code SUBCHANFLOW (SCF, developed by KIT, Germany) for a full-core VVER model. For such purpose, a recently refurbished master-slave coupling scheme is used within a High Performance Computing architecture. A full-core benchmark for a VVER-1000 that provides experimental data is considered, where the first burnup step (i.e. fresh core at hot-full rated power state) is calculated. For such purpose a detailed (i.e. pin-by-pin) coupled Serpent-SCF model is developed, including a simplified equilibrium xenon distribution (i.e. by fuel assembly). Comparisons with main global reported results are presented and briefly discussed, together with a raw estimation of resources requirements and a brief demonstration of the inherent capabilities of the proposed approach. The results presented here provide valuable insights and pave the way to tackle the final goals of the on-going high-fidelity project.

Development of high-fidelity neutron transport code STREAM

2021 ◽  
pp. 107915
Author(s):  
Sooyoung Choi ◽  
Wonkyeong Kim ◽  
Jiwon Choe ◽  
Woonghee Lee ◽  
Hanjoo Kim ◽  
...  
Keyword(s):  
Neutron Transport ◽  
High Fidelity ◽  
Transport Code

scholarly journals Numerical Approximation for Fractional Neutron Transport Equation

2021 ◽  
Vol 2021 ◽  
pp. 1-14
Author(s):  
Zhengang Zhao ◽  
Yunying Zheng
Keyword(s):  
Transport Equation ◽  
Nuclear Reactor ◽  
Neutron Transport ◽  
Transport Processes ◽  
Neutron Transport Equation ◽  
Spatial Direction ◽  
Fully Discrete ◽  
Discrete Methods ◽  
Fifth Order ◽  
Different Order

Fractional neutron transport equation reflects the anomalous transport processes in nuclear reactor. In this paper, we will construct the fully discrete methods for this type of fractional equation with Riesz derivative, where the generalized WENO5 scheme is used in spatial direction and Runge–Kutta schemes are adopted in temporal direction. The linear stabilities of the generalized WENO5 schemes with different stages and different order ERK are discussed detailed. Numerical examples show the combinations of forward Euler/two-stage, second-order ERK and WENO5 are unstable and the three-stage, third-order ERK method with generalized WENO5 is stable and can maintain sharp transitions for discontinuous problem, and its convergence reaches fifth order for smooth boundary condition.

A POD reduced-order model for resolving the neutron transport problems of nuclear reactor

2020 ◽  
Vol 149 ◽  
pp. 107799
Author(s):  
Yue Sun ◽  
Junhe Yang ◽  
Yahui Wang ◽  
Zhuo Li ◽  
Yu Ma
Keyword(s):  
Nuclear Reactor ◽  
Neutron Transport ◽  
Reduced Order Model ◽  
Order Model ◽  
Reduced Order ◽  
Transport Problems

scholarly journals Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core

2020 ◽  
Vol 2020 ◽  
pp. 1-12 ◽  
Author(s):  
Linrong Ye ◽  
Mingjun Wang ◽  
Xin’an Wang ◽  
Jian Deng ◽  
Yan Xiang ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Coupling Method ◽  
High Fidelity ◽  
Monte Carlo Code ◽  
Coupling Analysis ◽  
Fuel Reactor ◽  
Nuclear Reactor Core ◽  
Type Fuel ◽  
Plate Type

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

scholarly journals A High-Fidelity Approach for the Simulation of Flow-Induced Vibration

2016 ◽  
Author(s):  
Elia Merzari ◽  
Jerome Solberg ◽  
Paul Fischer ◽  
Robert M. Ferencz
Keyword(s):  
Reactor Core ◽  
Spectral Element ◽  
Structural Mechanics ◽  
Nuclear Industry ◽  
High Fidelity ◽  
Flow Induced Vibration ◽  
Strong Basis ◽  
Flow Past A Cylinder ◽  
Flow Induced Vibrations ◽  
Computational Resources

Flow-induced vibration (FIV) is a widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occur. Given the importance of reliable components in the nuclear industry, flow-induced vibrations have long been a major concern in the safety and operation of nuclear reactors. In particular, nuclear fuel and steam generators have been known to suffer from flow-induced vibrations and related failures. Over the past five years, the Nuclear Energy Advanced Modeling and Simulation program has developed the integrated multiphysics code suite SHARP. The goal of developing such a tool is to perform multiphysics modeling of the components inside a reactor core, the full reactor core or portions of it, and be able to achieve that with various levels of fidelity. This flexibility allows users to select the appropriate level of fidelity for their computational resources and design constraints. In particular SHARP contains high-fidelity single-physics codes for structural mechanics and fluid mechanics calculations: the structural mechanics implicit code Diablo and the computational fluid dynamics spectral element code Nek5000. Both codes are state-of-the-art. highly scalable (up to millions of processors in the case of Nek5000) tools that have been extensively validated. These tools form a strong basis on which to build an FIV modeling capability. This work discusses in detail the implementation of a fluid-structure interaction methodology in SHARP for simulating flow-induced viration based on the coupling between Diablo and Nek5000. Initial verification and validation efforts are also discussed, with a focus on standard benchmark cases: the flow past a cylinder, the Turek benchmark, and the flow in a Coriolis flow meter.

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