• Korean Chemical Engineering Research
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• v.59 no.3
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• pp.467-479
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• 2021

Abstract: A kinetic model was developed for the dehydration of ethanol to ethylene based on two parallel reaction pathways. Kinetic parameters were estimated by fitting experimental data of powder catalysts in a lab-scale test, and the effectiveness factor was determined using data from pellet-type catalysts in bench-scale experiments. The developed model was used to design a multitubular fixed-bed reactor (MTR) and an adiabatic reactor (AR) at a 10 ton per day scale. The two different reactor types resulted in different process configurations: the MTR consumed the ethanol completely and did not produce the reaction intermediate, diethyl ether (DEE), resulting in simple separation trains at the expense of high equipment cost for the reactor, whereas the AR required azeotropic distillation and cryogenic distillation to recycle the unreacted ethanol and to separate the undesired DEE, respectively. Quantitative analysis based on the equipment and annual energy costs showed that, despite high equipment cost of the reactor, the MTR process had the advantages of high productivity and simple separation trains, whereas the use of additional separation trains in the AR process increased both the total equipment cost and the annual energy cost per unit production rate.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• Korean Chemical Engineering Research
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• v.46 no.5
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• pp.931-937
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• 2008

Rigorous multiscale modelling and simulation of the MTR for WGSR was carried out to accurately predict the behavior of process variables and the reactor performance. The MTR consists of 4 fixed bed tube reactors packed with heterogeneous catalysts, as well as surrounding shell part for the cooling purpose. Considering that fluid flow field and reaction kinetics give a great influence on the reactor performance, employing multiscale methodology encompassing Computational Fluid Dynamics (CFD) and process modeling was natural and, in a sense, inevitable conclusion. Inlet and outlet temperature of the reactant fluid at the tube side was $345^{\circ}C$ and $390^{\circ}C$, respectively and the CO conversion at the exit of the tube side with these conditions approached to about 0.89. At the shell side, the inlet and outlet temperature of the cooling fluid, which flows counter-currently to tube flow, was $190^{\circ}C$ and $240^{\circ}C$. From this heat exchange, the energy saving was achieved for the flow at shell side and temperature of the tube side was properly controlled to obtain high CO conversion. The simulation results from this research were accurately comparable to the experimental data from various papers.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1266-1276
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• 2018

In this paper, three-dimensional neutronic calculations were performed in order to calculate the dependency of CRW on the temperature of fuel and moderator and the moderator void. Calculations were performed using the known MTR_PC computer codes in the core configuration 61 of TRR. The dependency of CRW on the fuel temperature in the range of $20-340^{\circ}C$ and the moderator temperature of each control rods were studied. Based on the positions of the control rods, the calculations were performed in three different cases, named case A, B and C. By the results, the worth of each control rods increases by increasing of the coolant temperature in all methods, however, the total CRW is somewhat independent of the fuel temperature. In addition, the results showed that the variation of CRW versus density depends on the positions of the control rods and the most change in CRW in the coolant temperature, $20-100^{\circ}C$ (279 pcm), belongs to SR4. Finally the effect of void on CRW was studied for different void fraction in coolant. The most worth change is about $2 for 40% void fraction related to SR1 and SR3 in case B. For 40% void fraction, the total CRW increases about $7.5, $6 and $7 in cases, A, B and C, respectively.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.700-709
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• 2017

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density ($U_3Si_2-Al$) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1017-1023
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• 2018

The measurement and calculation of neutronic parameters in nuclear research reactors has an important influence on control and safety of the nuclear reactor. The power peaking factors, reactivity coefficients and kinetic parameters are the most important neutronic parameter for determining the state of the reactor. The position of the control shim safety rods in the core configuration affects these parameters. The main purpose of this work is to use the MTR_PC package to evaluate the effect of the partially insertion of the control rod on the neutronic parameters at the operating core of the Tehran Research Reactor. The simulation results show that by increasing the insertion of control rods (bank) in the core, the absolute values of power peaking factor, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time decreased. In addition, the results show that the changes of moderator temperature coefficients value versus the control rods positions are very significant. The average value of moderator temperature coefficients increase about 98% in the range of 0-70% insertion of control rods.