Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

Reis, Paulo Rebelles; Costa, Antonella Lombardi; Pereira, Cláubia; Silva, Clarysson A. M.; Veloso, Maria Auxiliadora F.; Mesquita, Amir Zacarías

doi:10.1016/j.nucengdes.2011.10.022

Cited by 10 publications

(8 citation statements)

References 6 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…A better result for the transient condition was obtained when the same model was modified by dividing the reactor pool into two regions and adding a model of cross flow between both regions . 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).…”

Section: Introductionmentioning

confidence: 99%

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

Antariksawan¹,

Umar²,

Widodo³

et al. 2017

IJTech

View full text Add to dashboard Cite

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.

show abstract

Section: Introductionmentioning

confidence: 99%

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

Antariksawan¹,

Umar²,

Widodo³

et al. 2017

IJTech

View full text Add to dashboard Cite

show abstract

“…RELAP5 nodalization for the IPR-R1, verified in previous work (with 91 THC) [8], was utilized in this simulation. The core model has 91 hydrodynamic channels (see Figure 2(b)).…”

Section: Ipr-r1 Nodalizationmentioning

confidence: 99%

“…The core model has 91 hydrodynamic channels (see Figure 2(b)). The complete reactor nodalization in the RELAP5 is presented in Figure 3 [8][9][10]. To simulate the cross flow among the THCs a total of 186 single junctions were used, interconnecting selected volumes at different high positions of the nodalized channels.…”

Section: Ipr-r1 Nodalizationmentioning

confidence: 99%

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Reis

Costa

Pereira

et al. 2015

Science and Technology of Nuclear Installations

Self Cite

View full text Add to dashboard Cite

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.

show abstract

“…Sensitivity analysis including systematic variations in code input variables or modelling parameters, must be used to help identify the important parameters necessary for an accident analysis by ranking the influence of accident phenomena or to bound the overall Safety Studies and General Simulations of Research Reactors Using Nuclear Codes 29 results of the analysis. Results of experiments can also be used to identify important parameters (Reis et al, 2011).…”

Section: Verification and Validation Of Codesmentioning

confidence: 99%

Nuclear Power - System Simulations and Operation

Tsvetkov¹

2011

View full text Add to dashboard Cite

Statements and opinions expressed in the chapters are these of the individual contributors and not necessarily those of the editors or publisher. No responsibility is accepted for the accuracy of information contained in the published articles. The publisher assumes no responsibility for any damage or injury to persons or property arising out of the use of any materials, instructions, methods or ideas contained in the book.

show abstract

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

Cited by 10 publications

References 6 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

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Nuclear Power - System Simulations and Operation

Contact Info

Product

Resources

About