The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

Abstract

The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature

Analysis of Loss of Flow Events on Brazilian Multipurpose Reactor Using the Relap5 Code

“…The thermal hydraulic system code RELAP5 has been developed to best estimate transient simulation of light water power reactor systems during postulated accidents. Recent works have demonstrated that the code can be also used with good predictions for thermal hydraulic analysis of research reactors as it can be verified in the present literature [1][2][3][4][5][6][7][8].…”

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

“…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.…”

Thermal-hydraulic analysis under partial loss of flow accident hypothesis of a plate-type fuel surrounded by two water channels using RELAP5 code

“…In recent works, the code was used to thermal-hydraulic analysis of RRs. [13][14][15][16][17] According to RELAP5 manual, a table lookup method developed by Groeneveld, Cheng, and Doan is used for the prediction of the CHF. Code users also have the option of using a CHF correlation developed by the Czech Republic.…”