Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

Imke, Uwe; Sánchez, V.

doi:10.1155/2012/465059

Cited by 60 publications

(14 citation statements)

References 5 publications

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“…In this section work we describe the COTENP code to calculate the thermal-hydraulic parameters of nuclear reactors with plate type fuel elements. To represent the tridimensional behavior of the core we combine the subchannel approach and the cascade method in a two stage scheme as shown in Figure 1 [2,5,13,27]. In the first stage the core is divided into subchannels with size equivalent to a fuel assembly.…”

Section: Methods For the Cotenp Codementioning

confidence: 99%

“…Most subchannel analysis codes were developed for rod type fuels and fewer are dedicated to plate type geometry. Many belong to the COBRA family evolving from the first ones developed in the 1970s [11][12][13][14][15][16][17][18][19]. The COBRA-3C/RERTR code is an example of such developments applied for plate type fuel research and test reactors with low pressure [17].…”

Section: Introductionmentioning

confidence: 99%

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Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Castellanos-Gonzalez

Moreira

Carajilescov

2018

Science and Technology of Nuclear Installations

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This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and multipurpose reactors. To solve the conservation equations for mass, momentum, and energy, we adopt the subchannel and control volume methods based on fuel assembly geometric data and thermal-hydraulic conditions. We consider the chain or cascade method in two steps to facilitate the analysis of whole core. In the first step, we divide the core into channels with dimensions equivalent to that of the fuel assembly and identify the assembly with largest enthalpy rise as the hot assembly. In the second step, we divide the hot fuel assembly into subchannels with size equivalent to one actual coolant channel and similarly identify the hot subchannel. The code utilizes the homogenous equilibrium model for two-phase flow treatment and the balanced drop pressure approach for the flow rate determination. The code results include detailed information such as core pressure drop, mass flow rate distribution, coolant, cladding and centerline fuel temperatures, coolant quality, local heat flux, and results regarding onset of nucleate boiling and departure of nucleate boiling. To validate the COTENP code, we considered experimental data from the Brazilian IEA-R1 research reactor and calculated data from the Chinese CARR multipurpose reactor. The mean relative discrepancies for the coolant distribution were below 5%, for the coolant velocity were 1.5%, and for the pressure drop were below 10.7%. The latter discrepancy can be partially justified due to lack of information to adequately model the IEA-R1 experiment and CARR reactor. The results show that the COTENP code is sufficiently accurate to perform steady-state thermal-hydraulic design analyses for reactors with plate type fuel assemblies.

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Section: Methods For the Cotenp Codementioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Castellanos-Gonzalez

Moreira

Carajilescov

2018

Science and Technology of Nuclear Installations

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“…As computational power becomes more easily accessible, the modelling industry is likely to incorporate these methods more. Recent developments have seen a short transient being performed in 3D by coupling Serpent [25], a Monte Carlo neutron transport solver and SUBCHANFLOW [26], a subchannel code this project produced a pin by pin power profile of a small core during a reactivity control insertion [27]. The computational power required was 1000 cores, which is currently seen as inaccessible to most industrial institutions; however, two decades ago, this achievement would have been thought to be impossible.…”

Section: Countrymentioning

confidence: 99%

A Comparison of Advanced Boiling Water Reactor Simulations between Serpent/CTF and Polaris/DYN3D: Steady State Operational Characteristics and Burnup Evolution

et al. 2021

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High fidelity modelling for nuclear power plant analysis is becoming more common due to advances in modelling software and the availability of high-performance computers. However, to design, develop and regulate new light water nuclear reactors there are, up until now, limited requirements for high fidelity methods due to the already well-established computational methods already being widely accepted. This article explores the additional detail which can be obtained when using high fidelity methods through Monte Carlo/Sub-channel analysis compared to industrial methods of cross-section/nodal analysis using the Advanced Boiling Water Reactor as a case study. This case study was chosen due to the challenges in modelling two phase flow and the high levels of heterogeneity within the fuel assembly design. The article investigates how to implement such an approach, from a bottom-up procedure, by analysing each stage of the modelling process.

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“…The ability to integrate sub-channel and system code in one software product [3], as well as relatively low requirements for computer resources [4], are the advantages of this approach. The main disadvantage of this approach is the need to determine the closing relations, in particular, the coefficients of interchannel mixing ratio, which directly depend on the corresponding geometry and flow regime [3][4][5][6]. These relations are not directly derived from the three-dimensional equations for the flow and should be determined mainly experimentally for similar conditions.…”

Section: Introductionmentioning

confidence: 99%

Application of an Integral Turbulence Model to Close the Model of an Anisotropic Porous Body as Applied to Rod Structures

Vlasov

Merinov

2022

Fluids

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In practice, often devices are ordered rod structures consisting of a large number of rods. Heat exchangers, fuel assemblies of nuclear reactors, and their cores in the case of using caseless assemblies are examples of such devices. Simulation of heat and mass transfer processes in such devices in porous-body approximation can significantly reduce the required resources compared to computational fluid dynamics (CFD) approaches. The paper describes an integral turbulence model developed for defining anisotropic model parameters of a porous body. The parameters of the integral turbulence model were determined by numerical simulations for assemblies of smooth rods, assemblies with spacer grids, and wire-wrapped fuel assemblies. The results of modeling the flow of a liquid metal coolant in an experimental fuel assembly with local blocking of its flow section in anisotropic porous-body approximation using an integral turbulence model are described. The possibility of using the model of an anisotropic porous body with the integral model of turbulence to describe thermal-hydraulic processes during fluid flow in rod structures is confirmed.

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Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

Cited by 60 publications

References 5 publications

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

A Comparison of Advanced Boiling Water Reactor Simulations between Serpent/CTF and Polaris/DYN3D: Steady State Operational Characteristics and Burnup Evolution

Application of an Integral Turbulence Model to Close the Model of an Anisotropic Porous Body as Applied to Rod Structures

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