Development of a thermal–hydraulic analysis code for research reactors with plate fuels

Lu, Qilin; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.anucene.2008.11.038

Cited by 36 publications

(8 citation statements)

References 13 publications

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“…For the open pool research reactors configuration, the probability of a blockage in the core upper plenum zone is higher than a blockage from the bottom [11]. This transient situation may be caused, for example, by swelling of the fuel or fall of some material in the reactor pool leading to blockage of one or more channels [12]. A fuel channel blockage event has different characteristics depending on the flow direction.…”

Section: Thcs Blockage Transientmentioning

confidence: 99%

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Reis

Costa

Pereira

et al. 2015

Science and Technology of Nuclear Installations

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Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

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Section: Thcs Blockage Transientmentioning

confidence: 99%

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Reis

Costa

Pereira

et al. 2015

Science and Technology of Nuclear Installations

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“…Their simulation showed that the integrity of the fuel was conserved even for a totally blocked channel case owing to cooling from the adjacent channels. Lu et al (2009b) also studied a single channel blockage transient with their in-house homogeneous two phase flow analysis code THAC-PRR where the reactivity feedback was taken into account. The calculations showed that a blockage lead to a redistribution of the coolant flows into the adjacent channels but the increase was minor.…”

Section: Introductionmentioning

confidence: 99%

Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

Son

Yang

Park

et al. 2015

Annals of Nuclear Energy

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“…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%

“…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%

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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Development of a thermal–hydraulic analysis code for research reactors with plate fuels

Cited by 36 publications

References 13 publications

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

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

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