International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Hainoun, A.; Doval, A.; Umbehaun, Pedro E.; Chatzidakis, Stylianos; Ghazi, N.; Park, S.; Mladin, M.; Shokr, A. M.

doi:10.1016/j.nucengdes.2014.06.041

Cited by 29 publications

(22 citation statements)

References 8 publications

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“…The thermal-hydraulic variables of subchannel analyses of plate type fuel reactors found in the literature are coolant, cladding, and fuel axial temperature distributions, fuel assembly flow rate distribution, pressure drops across the reactor core, mean coolant velocity in subchannels, and departure from nucleate boiling results. Below we describe the IEA-R1 research reactor and the CARR multipurpose reactor and their reported subchannel analysis data [7,22,26,28,34] and the approach taken in this work to carry out the COTENP code validation. Figure 6 shows the top view of configuration 247 and more information about the IEA-R1 reactor can be found in Appendix A.…”

Section: Validation Approach and Datamentioning

confidence: 99%

“…Reference [22] reports total core pressure drop measured with the dummy but does not present details such as exact location of the measurements and their experimental error. Coolant velocity (ΔP assembly) IFA location (see Figure 7)…”

Section: Validation Approach and Datamentioning

confidence: 99%

“…The COBRA-3C/RERTR code is an example of such developments applied for plate type fuel research and test reactors with low pressure [17]. The literature reports other plate type thermal-hydraulic subchannel codes [18,19] as well as the use of thermal-hydraulic system codes to perform plate type fuel subchannel analysis [20][21][22][23]. Most of the subchannel computer codes for plate type reactors reported in the literature are not for general use but dedicated for specific parameter verifications of new research, test, or multipurpose reactors [17][18][19]24].…”

Section: Introductionmentioning

confidence: 99%

“…The code is based on current literature of subchannel analysis models and its accuracy is validated against experimental and calculated data reported for two reactors with plate type fuels. The first set of data belongs to the Research Reactor Benchmark Database from the International Atomic Energy Agency, more specifically, data from an Instrumented Fuel Assembly (IFA) of the 5 MW IEA-R1 research reactor [21,22,26]. The second set of data comes from calculation results for the 50 MW CARR multipurpose reactor [7].…”

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: Validation Approach and Datamentioning

confidence: 99%

Section: Validation Approach and Datamentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

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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“…Many of these computational tools or codes are initially developed for a nuclear power reactor, such as RELAP5, ATHLET and CATHARE codes. However, several study have been done to assess the applicability of these codes for different type of research reactors, especially MTR type [3][4][5][6], TRIGA type [7], and also for other nuclear test facilities [8,9].…”

Section: Introductionmentioning

confidence: 99%

Parametric Study of Loca in Triga-2000 Using Relap5/Scdap Code

2017

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PARAMETRIC STUDY OF LOCA IN TRIGA-2000 USING RELAP5/SCDAP CODE.A postulated loss of coolant accident (LOCA) shall be analyzed to assure the safety of a research reactor. The analysis of such accident could be performed using best estimate thermal-hydraulic codes, such as RELAP5. This study focuses on analysis of LOCA in TRIGA-2000 due to pipe and beam tube break. The objective is to understand the effect of break size and the actuating time of the emergency core cooling system (ECCS) on the accident consequences and to assess the safety of the reactor. The analysis is performed using RELAP/SCDAPSIM codes. Three different break size and actuating time were studied. The results confirmed that the larger break size, the faster coolant blow down. But, the siphon break holes could prevent the core from risk of dry out due to siphoning effect in case of pipe break. In case of beam tube rupture, the ECCS is able to delay the fuel temperature increased where the late actuation of the ECCS could delay longer. It could be concluded that the safety of the reactor is kept during LOCA throughout the duration time studied. However, to assure the integrity of the fuel for the long term, the cooling system after ECCS last should be considered.

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Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool

Khater

El-Morshedy

Abdelmaksoud

2015

Annals of Nuclear Energy

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International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Cited by 29 publications

References 8 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

Parametric Study of Loca in Triga-2000 Using Relap5/Scdap Code

Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool

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