Loss of coolant accident analyses on Tehran research reactor by RELAP5/MOD3.2 code

Hedayat, Afshin; Davilu, Hadi; Jafari, Jalil

doi:10.1016/j.pnucene.2007.07.009

Cited by 48 publications

(8 citation statements)

References 3 publications

(2 reference statements)

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“…The number of channels in the core was also found to be important to refine the calculation results (Reis et al, 2012). Other studies assessed several thermalhydraulic codes, including RELAP5, against an MTR-type reactor for a steady state condition, loss of flow, and loss of coolant transient (Abdelrazek et al, 2014;Chatzidakis et al, 2014;Chatzidakis et al, 2013;Hedayat et al, 2007;Karimpour & Esteki, 2015). In those studies, the RELAP5 code was generally able to simulate the steady state condition with good agreement but with a higher discrepancy during the transient considered in the study.…”

Section: Introductionmentioning

confidence: 99%

TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4

Antariksawan¹,

Umar²,

Widodo³

et al. 2017

IJTech

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Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes. In this paper, the applicability of the RELAP/SCDAPSIM/MOD3.4 thermalhydraulic code to one Indonesian research reactor, which is named TRIGA-2000, is performed. The aim is to validate the model and use the model to analyze the thermal-hydraulic characteristics of TRIGA-2000 for main transient events considered in the Safety Analysis Report. The validation was done by comparing the calculation results with experimental data mainly in steady state conditions. The comparison of calculation results with the measurement data showed good agreement with little discrepancies. Based on these results, simulations for thermal-hydraulic analyses were performed for loss of coolant transients. The calculation results also properly depicted the physic of the thermal-hydraulic phenomena following the loss of coolant transients. These results showed the adequacy of the model. It could be shown that the engineered safety features of TRIGA-2000 play an important role in keeping the reactor safe from the risk of postulated loss of a coolant accident.

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Section: Introductionmentioning

confidence: 99%

TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4

Antariksawan¹,

Umar²,

Widodo³

et al. 2017

IJTech

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“…The coupling method predicts the thermal hydraulic behavior of the reactor in detail and evaluate the thermal feedback effects and also to simulate accident scenarios [7].This techniques incorporates to simulates transient which involves symmetric core spatial power distribution and strong feedback effects between neutronics and thermal hydraulic .It deals with reactivity initiated accidents with local core effects, coupled reactor core and plant dynamic interactions [3].The postulated accident that can be simulate safely includes control rod with drawl [8],loss of coolant accident [9],main steam line break accidents etc because during the accidents, deformation in the radial and axial power distribution can occurs in core [10,11].…”

Section: Neutronic Modelmentioning

confidence: 99%

Neutronics and Thermal Hydraulic Coupling Methods for the Nuclear Reactor Core

Ud-Din¹,

Peng²,

Zubair³

2011

2011 Asia-Pacific Power and Energy Engineering Conference

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In this paper an overview of the coupling methodologies between reactor physics (Neutronics) and thermal hydraulic have been studied by taken into account the previous and ongoing research. The current study includes the neutronics and thermal hydraulic model and the coupling methodology between the two phenomenons which is inter related with computational approach. Since the coupling involves some computer codes which exchanges the data between the neutronics and thermal hydraulic so in this paper possible computer codes are also have been listed with their categories. Finally the advantages of the coupling are described with some suggestions and conclusion is also given.

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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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Loss of coolant accident analyses on Tehran research reactor by RELAP5/MOD3.2 code

Cited by 48 publications

References 3 publications

TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4

TRIGA-2000 Research Reactor Thermal-hydraulic Analysis using RELAP/SCDAPSIM/MOD3.4

Neutronics and Thermal Hydraulic Coupling Methods for the Nuclear Reactor Core

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

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