Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

Reis, Paulo Rebelles; Costa, Antonella Lombardi; Pereira, Cláubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Zacarías; Soares, Humberto V.; Barros, Graiciany de Paula

doi:10.1016/j.anucene.2010.05.013

Cited by 30 publications

(13 citation statements)

References 8 publications

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“…The RELAP5 model developed for IPR-R1 reactor showed a good agreement with the experimental data for 50 kW steady state, but not for loss of flow transient condition [10]. The improvement of the model by dividing the reactor pool into two regions with a cross flow model between both regions showed a better transient calculation result [7]. A sensitivity study on the core model of IPR-R1 showed that better coolant temperature prediction was obtained with increasing number of core channels and cross flow models [11].…”

Section: Introductionmentioning

confidence: 56%

“…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%

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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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Section: Introductionmentioning

confidence: 56%

Section: Introductionmentioning

confidence: 99%

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

2017

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“…Currently, with the calculation codes such as: APROS, RELAP, CATHARE, and ATHLET it has become possible to provide the thermal hydraulic behavior of the facilities and to reproduce the physical phenomena occurring during normal and accidental operation (Nematollahi and Zare, 2008;Reis et al, 2010). These computer codes are the result of scientific research conducted for several decades by international collaboration in the field of safety and design of nuclear power plants.…”

Section: Introductionmentioning

confidence: 99%

Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

Cheridi

Chaker

Loubar

2016

Annals of Nuclear Energy

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“…For example, a subcooled boiling model of upward vertical flow consistent with phenomenological observations of the subcooled flow boiling mechanisms was proposed to extend the range of applicability of the RELAP5 code to low pressures (Končar and Mavko, 2003). Therefore, recent works as, for example (Antariksawan et al, 2005;Khedr et al, 2005;Marcum et al, 2010;Reis et al, 2010), have been performed to investigate the applicability of the RELAP5 code to research reactors operating conditions (TRIGA 2000, MTR, Oregon State TRIGA, IPR-R1 TRIGA), respectively. Application of a model for the IPR-R1 TRIGA using the RELAP5 code is detailed in the Annex B.…”

Section: Thermal Hydraulic Modelingmentioning

confidence: 99%