Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes

High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.

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Survey of Coupling Schemes in Traditional Coupled Neutronics and Thermal-Hydraulics Codes

Volume 6B: Energy ◽

10.1115/imece2015-52990 ◽

2015 ◽

Author(s):

Sabahattin Akbas ◽

Victor Martinez-Quiroga ◽

Fatih Aydogan ◽

Abderrafi M. Ougouag ◽

Chris Allison

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Nuclear Reactor ◽

Primary System ◽

Thermal Hydraulics ◽

Neutron Kinetics ◽

Coupled Codes ◽

Physical Phenomena ◽

Selection Of

The design and the analysis of nuclear power plants (NPPs) require computational codes to predict the behavior of the NPP nuclear components and other systems (i.e., reactor core, primary coolant system, emergency core cooling system, etc.). Coupled calculations are essential to the conduct of deterministic safety assessments. Inasmuch as the physical phenomena that govern the performance of a nuclear reactor are always present simultaneously, ideally computational modeling of a nuclear reactor should include coupled codes that represent all of the active physical phenomena. Such multi-physics codes are under development at several institutions and are expected to become operational in the future. However, in the interim, integrated codes that incorporate modeling capabilities for two to three physical phenomena will remain useful. For example, in the conduct of safety analyses, of paramount importance are codes that couple neutronics and thermal-hydraulics, especially transient codes. Other code systems of importance to safety analyses are those that couple primary system thermal-hydraulics to fission product chemistry, neutronics to fuel performance, containment behavior and structural mechanics to thermal-hydraulics, etc. This paper surveys the methods used traditionally in the coupling of neutronic and thermal-hydraulics codes. The neutron kinetics codes are used for computing the space-time evolution of the neutron flux and, hence, of the power distribution. The thermal-hydraulics codes, which compute mass, momentum and energy transfers, model the coolant flow and the temperature distribution. These codes can be used to compute the neutronic behavior and the thermal-hydraulic states separately. However, the need to account with fidelity for the dynamic feedback between the two sets of properties (via temperature and density effects on the cross section inputs into the neutronics codes) and the requirement to model realistically the transient response of nuclear power plants and to assess the associated emergency systems and procedures imply the necessity of modeling the neutronic and thermal-hydraulics simultaneously within a coupled code system. The focus of this paper is a comparison of the methods by which the coupling between neutron kinetics and thermal-hydraulics treatments has been traditionally achieved in various code systems. As discussed in the last section, the modern approaches to multi-physics code development are beyond the scope of this paper. From the field of the most commonly used coupled neutron kinetic-thermal-hydraulics codes, this study selected for comparison the coupled codes RELAP5-3D (NESTLE), TRACE/PARCS, RELAP5/PARCS, ATHLET/DYN3D, RELAP5/SCDAPSIM/MOD4.0/NESTLE. The choice was inspired by how widespread the use of the codes is, but was limited by time availability. Thus, the selection of codes is not to be construed as exhaustive, nor is there any implication of priority about the methods used by the various codes. These codes were developed by a variety of institutions (universities, research centers, and laboratories) geographically located away from each other. Each of the research group that developed these coupled code systems used its own combination of initial codes as well as different methods and assumptions in the coupling process. For instance, all these neutron kinetics codes solve the few-groups neutron diffusion equations. However, the data they use may be based on different lattice physics codes. The neutronics solvers may use different methods, ranging from point kinetics method (in some versions of RELAP5) to nodal expansion methods (NEM), to semi-analytic nodal methods, to the analytic nodal method (ANM). Similarly, the thermal-hydraulics codes use several different approaches: different number of coolant fields, homogenous equilibrium model, separate flow model, different numbers of conservation equations, etc. Therefore, not only the physical models but also the assumptions of the coupled codes and coupling techniques vary significantly. This paper compares coupled codes qualitatively and quantitatively. The results of this study are being used both to guide the selection of appropriate coupled codes and to identify further developments into coupled codes.

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A Simplified Two-Group Multipoint Kinetics Model

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4046254 ◽

2020 ◽

Vol 7 (1) ◽

Author(s):

Santosh K. Pradhan ◽

K. Obaidurrahman ◽

Kannan N. Iyer

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Three Dimensional ◽

Reactor Core ◽

Coupled System ◽

Kinetics Model ◽

Primary Circuit ◽

Thermal Hydraulics ◽

Time Step ◽

Large Power

Abstract Detailed multiphysics modeling of nuclear power plants has become a necessity in the era of best-estimate analysis. For a number of transients with strong coupling between the neutronics in the reactor core and the fluid-dynamics in the primary circuit and overall heat transfer, it is required to carry out coupled system thermal hydraulics and core three-dimensional (3D) neutronics analysis. Point kinetics approach in the system thermal-hydraulics (TH) code RELAP5 limits its use for many reactivity-induced transients, which involve asymmetric core behavior. In a recent development, a simplified multipoint kinetics model has been coupled with system TH code RELAP5 to circumvent its inadequacy for the analysis of reactivity-induced transients involving asymmetric core behavior. The objective of this paper is to validate the simplified multipoint kinetics model against an asymmetric fast transient benchmark problem in a large power reactor. Time-step and nodalization sensitivity studies have been performed. It is demonstrated that the multipoint kinetics model results are in good agreement with the benchmark, advocating its applicability.

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Hydrogen Deflagration Analysis in THAI Experiments Using MELCOR Code

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4048006 ◽

2020 ◽

Author(s):

Alain Flores y Flores ◽

Guido Mazzini

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Control Volume ◽

Critical Issue ◽

Severe Accident ◽

Test Facility ◽

Research Centre ◽

Time Step ◽

Volume Number ◽

Fukushima Daiichi

Abstract In order to develop an appropriate knowledge to support the SUJB (State Office of Nuclear Safety), the CVR (Research Centre Re), in collaboration with SURO (National Radiation Protection Institute) is developing a methodology to simulate nuclear power plants under accidental conditions. A particular effort is focused in the severe accident phenomenology where hydrogen deflagration carries a critical issue for the containment integrity, such as Fukushima Daiichi accident. For this purpose, THAI (Thermal-hydraulics, hydrogen, aerosol and iodine) experimental campaigns are chosen due to the several tests involved in different conditions. THAI containment test facility is used to open questions concerning the behaviour of hydrogen, iodine and aerosols in the containment of water-cooled reactors during severe accidents. The Fukushima Daiichi Accident demonstrates that the hydrogen deflagration could lead to a significant containment damage. For this reason, a particular attention is given to the hydrogen deflagration scenario. All simulations are prepared and modelled in MELCOR 2.1. The results obtained showed a strong influence related with some factors as: the nodalization pattern, control volume number (CV), flow paths number FP and time step. In order to assess the THAI model with the THAI final reports, a sensitivity analysis focused with those parameters was performed.

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Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type

Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors ◽

10.1115/icone2020-16326 ◽

2020 ◽

Author(s):

Eduard Usov ◽

Nikolay Pribaturin ◽

Vladimir Chukhno ◽

Ilya Klimonov ◽

Anton Butov ◽

...

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Liquid Metals ◽

Numerical Algorithms ◽

Severe Accident ◽

Theoretical Research ◽

Detailed Calculation ◽

Fuel Rods ◽

Fuel Rod ◽

The Core

Abstract Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

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HIGH-ORDER FINITE ELEMENTS FOR THE NEUTRON TRANSPORT EQUATION ON HONEYCOMB MESHES

EPJ Web of Conferences ◽

10.1051/epjconf/202124703002 ◽

2021 ◽

Vol 247 ◽

pp. 03002

Author(s):

Ansar Calloo ◽

Romain Le Tellier ◽

David Labeurthre

Keyword(s):

Finite Elements ◽

Transport Equation ◽

Neutron Transport ◽

Computational Cost ◽

Rational Functions ◽

High Order ◽

Severe Accident ◽

Honeycomb Lattice ◽

Spatial Problem ◽

Neutron Transport Equation

Presently, APOLLO3®/MINARETsolves the transport equation using the multigroup Sn method with discontinuous finite elements on triangular meshes with Lagrange polynomial bases. The goal of this work is to solve the spatial problem on hexagonal geometries in the context of honeycomb lattice reactors, without further refining the computational mesh. The idea here is to construct high-order basis functions on the hexagonal element in order to improve the trade-off between computational cost and accuracy, in particular for multiphysics simulations where, often, thermalhydraulic modelling requires only assembly-average cross-sections to be defined (e.g.severe accident of fast breeder reactors)i.e.the assemblies are assumed homogeneous. One approach to achieve this goal is through the use of generalised barycentric functions such as the Wachspress rational functions. This research endeavour deals with the application of Wachspress rational functions to the neutron transport equation for hexagonal geometries up to order 3. With this method, it is possible to decrease the number of spatial unknowns required for the same accuracy, and thus the computational burden for complex geometries, such as honeycomb lattices is reduced.

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KYLIN-V2.0 CODE CALCULATION ABILITY VERIFICATION BASED ON VERA BENCHMAR

EPJ Web of Conferences ◽

10.1051/epjconf/202124710020 ◽

2021 ◽

Vol 247 ◽

pp. 10020

Author(s):

Dongyong Wang ◽

Yingrui Yu ◽

Xingjie Peng ◽

Chenlin Wang ◽

Kun Liu ◽

...

Keyword(s):

Fuel Assembly ◽

Power Distribution ◽

Nuclear Power ◽

Neutron Transport ◽

Monte Carlo Code ◽

Neutron Transport Equation ◽

Control Rod ◽

Fuel Assemblies ◽

Fuel Pin ◽

Calculation Ability

Virtual Environmental for Reactor Analysis (VERA) benchmark was released by the Consortium for Advanced Simulation of Light water reactors (CASL) project in 2012. VERA benchmark includes more than ten problems at different levels, from 2D fuel pin case to 2D fuel assembly case to 3D core refuelling case, in addition, reference results and experimental measured data of some problems were provided by CASL. Fuel assemblies in VERA benchmark are various, including control rod assemblies, Pyrex assembly, IFBA assembly, WABA assembly and gadolinium poison assembly, and so on. In this paper, various fuel assembly models in the VERA benchmark have been built by using KYIIN-V2.0 code to verify its calculation ability from 2D fuel pin case to 2D fuel assembly case to 2D 3x3 fuel assembly case, and making a comparative analysis on the reference results in VERA benchmark, as well as the calculation results of the Monte Carlo code RMC. KYLIN-V2.0 is an advanced neutron transport lattice code developed by Nuclear Power Institute of China (NPIC). The subgroup resonance calculation method is used in KYIIN-V2.0 to obtain effective resonance selfshielding cross section, method of modular characteristics (MOC) is adopted to solve the neutron transport equation, and CRAM method and PPC method is adopted to solve the depletion equation. The numerical results show that KYLIN-V2.0 code has the reliable capability of direct heterogeneous calculation of 2D fuel assembly, and the effective multiplication factor, assembly power distribution, rod power distribution and control rod reactivity worths of various fuel assemblies that are calculated by KYLIN-V2.0 are in better agreement with the reference.

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Thermal/electric energy generation and CO2 production for greenhouse facilities

Power and Autonomous equipment ◽

10.32464/2618-8716-2019-2-3-141-151 ◽

2019 ◽

Vol 2 (3) ◽

pp. 141-151

Author(s):

O. E. Gnezdova ◽

E. S. Chugunkova

Keyword(s):

Carbon Dioxide ◽

Power Distribution ◽

Power Plants ◽

Electric Energy ◽

Electricity Consumption ◽

Vegetable Crops ◽

Carbon Dioxide Injection ◽

Electricity Rates ◽

Traditional Patterns ◽

Generation Power

Introduction: greenhouses need microclimate control systems to grow agricultural crops. The method of carbon dioxide injection, which is currently used by agricultural companies, causes particular problems. Co-generation power plants may boost the greenhouse efficiency, as they are capable of producing electric energy, heat and cold, as well as carbon dioxide designated for greenhouse plants.Methods: the co-authors provide their estimates of the future gas/electricity rates growth in the short term; they have made a breakdown of the costs of greenhouse products, and they have also compiled the diagrams describing electricity consumption in case of traditional and non-traditional patterns of power supply; they also provide a power distribution pattern typical for greenhouse businesses, as well as the structure and the principle of operation of a co-generation unit used by a greenhouse facility.Results and discussion: the co-authors highlight the strengths of co-generation units used by greenhouse facilities. They have also identified the biological features of carbon dioxide generation and consumption, and they have listed the consequences of using carbon dioxide to enrich vegetable crops.Conclusion: the co-authors have formulated the expediency of using co-generation power plants as part of power generation facilities that serve greenhouses.

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