Search

The present work was developed within the frame of the IAEA Coordinated Research Project 1496 “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. Three benchmark experiments were designed and performed to study the performance of the Egyptian Research Reactor ETRR-2. The experiments included steady state measurements as well as loss of flow transient (LOFT) and loss of power transient (SCRAM) conditions. The Code RELAP5/Mod3.4 was used to simulate the components of the ETRR-2 systems for the thermal hydraulic analysis of the reactor. The dimensions and elevations of the primary cooling components are based on real conditions (3-D configuration). RELAP5 results provided benchmark data which verified the experimental measurements taken from instrumentations installed for this experiment at several positions in the core. The predicted values for the coolant and clad surface temperature at different locations in the core showed a remarkable agreement with the experimental values, for both the steady state and transient conditions.

The present work is developed within the frame of the IAEA Coordinated Research Program 1496, “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal-hydraulic computational methods and tools for operation and safety analysis of research reactors”. The objective of this work is to test the credibility of RELAP5 code in simulating the research reactor behavior during transients. Where, the code’s results were compared to the experimental measurements taken from Instrumented (thermocouple) Fuel Elements (IFE), installed at several positions in the core. The research reactor used in this study is the RSG-GAS reactor, located in the Serpong in Indonesia. Both the steady state and the transient simulation results performed using RELAP5 code for, are presented in this work. The calculated values for the coolant and clad surface temperature, at different locations in the core, showed an agreement with the experimental values with a difference less than 7% for the steady state, and with a difference less than 10% for transient conditions. However, RELAP5 was unable to predict the measured value of the coolant output temperature after the natural convection begins especially after stagnation occurred. In this case, the difference between the code results and experimental data was about 23%.

A computation code THERMOHYD has been developed to perform the steady state thermal-hydraulic calculations of the MTR type research reactors fueled with plate-type-fuel, in the ranges of pressures and temperatures typical for MTR reactors. The code calculates the coolant and clad surface temperatures along the cooling channel, as well as the critical heat flux and provides values for the critical phenomena parameters (RDR, ONBR and DNBR ratios) for reactor cores. The code takes into account the variation of the physical properties of the coolant with temperature. It also accounts for the uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations which have been considered in the analysis. The THERMOHYD code was tested against the well known PARET code and proved to be accurate and more conservative in some aspects.

A steady state thermal-hydraulic analysis for the ETRR-2 Research Reactor has been performed using the computer code TERMIC. The analysis has included both, the verification of the core design as well as the performance of the equilibrium core. The uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations have been considered in the analysis. A methodology based on uncertainties propagation has been adopted for the evaluation of hot channel factors. The results of the thermal-hydraulic analysis show that the reactor can be operated at full power level with sufficient safety margins against the thermal-hydraulic critical phenomena.