Evaluation of RELAP5/MOD3 behavior against loss of flow experimental results from two research reactor facilities

Chatzidakis, Stylianos; Ikonomopoulos, A.; Ridikas, D.

doi:10.1016/j.nucengdes.2012.11.005

Cited by 17 publications

(8 citation statements)

References 14 publications

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“…As shown in Figure 10, the reactor core could be cooled by that natural circulation, and the fuel temperature could be maintained as low without significantly increasing the coolant temperature. The occurrence of natural circulation in the reactor pool was also found in another study (Chatzidakis et al, 2013). From this result, it could be seen that the siphon breaker holes were very important because they stopped the leakage.…”

Section: Steady Statesupporting

confidence: 62%

“…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%

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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: Steady Statesupporting

confidence: 62%

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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“…Since the Safety Analysis Report (SAR) ensured the safety of the fuel plates during the shutdown condition [4], an assumed failure of the shutdown system is made for this event. The event to be simulated will follow the scenario as described in the SAR as also already simulated through a benchmark activities [10,15]. The difference is that the simulation will focus directly in the cladding temperature of the 21 fuel plates as previously modelled.…”

Section: Cladding Temperatures Due To Loss Of Complete Flow With Reacmentioning

confidence: 99%

Relap5 Simulation for Severe Accident Analysis of RSG-Gas Reactor

Ekariansyah

Hastuti²,

Sudarmono³

2018

Tri Dasa Mega

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RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR.The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational chararacteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element.Keywords: loss of flow, blockage, fuel plate, RSG-GAS, RELAP5 ABSTRAK SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK) dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan

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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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Evaluation of RELAP5/MOD3 behavior against loss of flow experimental results from two research reactor facilities

Cited by 17 publications

References 14 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

Relap5 Simulation for Severe Accident Analysis of RSG-Gas Reactor

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

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