MW Reactor Research Articles

Biodiesel produced from waste cooking oil by microwave transesterification using inert aluminium foil (1%) as a heating promoter

The manufacture of biodiesel from waste cooking oil, using cost-effective and environmentally friendly processes, continues to attract attention. Microwave radiation (MW), which is easy to use and is not demanding on power supplies, can promote rapid chemical reactions; however, our understanding of the MW absorption that occurs with the oscillation of polar reactants is far from complete. To increase the interaction of MW with reactant, our approach used a MW reactor in conjunction with a transesterification process. A small amount of aluminium foil, which remains chemically inert in the reaction, was used as a MW heating promoter to accelerate transesterification without generating sparks in a process employing 700W MW irradiation at room temperature. The temperature of the MW reactor was measured by thermal image analysis, while thermogravimetric analysis (TA) and gas chromatography/mass spectroscopy were used to characterize the components and boiling points and of soybean oil and its fatty acid components. The optimal size of the aluminium foil to use was determined by studying the transesterification of soybean oil with catalytic NaOH in a closed glass bottle. The results indicated that the reaction time to give a conversion of 82% was shortened by approximately 60% by adding 1 wt% aluminium foil (3 × 3 mm2).

Electrified steam methane reforming of biogas for sustainable syngas manufacturing and next-generation of plant design: A pilot plant study

Electrification of the traditional steam methane reforming (SMR) technology for syngas manufacturing has a significant CO2 reduction potential and make it more feasible to operate in combination with carbon capture/utilization, especially if renewable electricity is used. This pilot plant study demonstrates the first operational experience with industrial-scale electrified steam methane reforming (eSMR) technology using biogas as sustainable carbon feedstock. Across an operating envelope spanning from combinations of 5 to 20 barg and 750 °C to 1000 °C, the eSMR produces syngas as expected from thermodynamics. However, without the thermal restrictions inherent in the SMR design, the eSMR can achieve higher temperatures, enables fast transient operation, with high stability and control, thereby increasing the overall performance and design flexibility. Initial thermal responses of the eSMR were tested, demonstrating fast start-up from an idle state to operating conditions, within 2.6 h, including heating from 630 °C to 900 °C. Furthermore, dynamic temperature control was also demonstrated with heating rates up to 330 °C/h. Experimental energy efficiency of the pilot reactor was quantified between 72 % and 80 %, with the residual being heat loss to the surroundings due to the relatively small scale. With further scale-up to ≥1 MW reactor capacity, efficiencies of ≥99 % are predicted with a specific electrical energy consumption of 1.0 kWh/Nm3 H2. Overall, the efficiency and operational flexibility are improved due to the direct electrical heating of the catalytic system. Combined with biogas as feedstock, this paves the way for attractive and competitive plant designs for sustainable and renewable production of chemicals and fuels.

H2 generation by the 10B(n,α)7Li reaction in high temperature water

H2 produced in water from the 10B(n,α)7Li fission reaction has been measured up to 300 °C. Thermal energy neutrons from the Rhode Island Nuclear Science Center's 2 MW reactor interact with boric acid-containing water in temperature-controlled high-pressure cells made from tubing of either titanium or zirconium alloy. After exposure for a minimum of 1 h, the solution sample is extracted and sparged with argon. The H2 entrained by the sparging gas is sampled with a small mass spectrometer. A small amount of sodium is included in the boric acid solution so that after sparging, samples can be collected for 24Na activation measurements in a gamma spectrometer to determine the neutron exposure and thus the total energy deposited in solution. The G-value (μmol/J) for H2 production is obtained for water at a pressure of 25 MPa, over a temperature range from 20 °C to 300 °C. The weak temperature dependence of this yield between 150 °C and 200 °C demonstrates that the bimolecular reaction of pairs of eaq− is a very minor source of H2 in high LET tracks.

Non-standard neutrino interactions in light mediator models at reactor experiments

Compared to other neutrino sources, the huge anti-neutrino fluxes at nuclear reactor based experiments empower us to derive stronger bounds on non-standard interactions of neutrinos with electrons mediated by light scalar/vector mediators. At neutrino energy around 200 keV reactor anti-neutrino flux is at least an order of magnitude larger compared to the solar flux. The atomic and crystal form factors of the detector materials related to the details of the atomic structure becomes relevant at this energy scale as the momentum transfers would be small. Non-standard neutrino-electron interaction mediated by light scalar/vector mediator arises naturally in many low-scale models. We also propose one such new model with a light scalar mediator. Here, we investigate the parameter space of such low-scale models in reactor based neutrino experiments with low threshold Ge and Si detectors, and find the prospect of probing/ruling out the relevant parameter space by finding the projected sensitivity at 90% confidence level by performing a χ2-analysis. We find that a detector capable of discriminating between electron recoil and nuclear recoil signal down to a very low threshold such as 5 eV placed in reactor based experiment would be able to probe a larger region in parameter space compared to the previously explored region. A Ge (Si) detector with 10 kg-yr exposure and 1 MW reactor anti-neutrino flux would be able to probe the scalar and vector mediators with masses below 1 keV for coupling products sqrt{g_{nu }{g}_e} ∼ 1 × 10−6 (9.5 × 10−7) and 1 × 10−7 (8 × 10−8), respectively.

A novel nonlinear radiative heat exchanger for molten-salt applications

One of the main difficulties associated with using molten salts as heat transfer fluids (HTFs) in practical applications such as in the solar and nuclear energy industries is the possibility of solidifying molten salts. Since the salts proposed for Generation-IV nuclear reactors (MSRs) have higher melting points, this is especially true for those reactors. The consequences of accidental freezing within the direct reactor auxiliary cooling system (DRACS) of MSRs could be catastrophic. DRACS is a passive safety system designed to remove decay heat from the reactor core during an emergency. The present study examines, as a case study, the freezing problem in the most vulnerable component of DRACS, namely the natural draft heat exchanger (NDHX). For this application, a unique and innovative heat exchanger design termed radiative heat exchanger (RHX) is proposed, with nonlinear heat transfer characteristics. Specifically, by suppressing convection and exploiting thermal radiation as the primary heat transfer mechanism, the RHX concept reduces the heat rejection/removal rate passively to allow safe operation when the molten salt coolant temperature drops and is near its melting point, while maintaining a high capacity for decay heat extraction in case the coolant temperature rises during an accident and the reactor becomes overheated; this moderation of the heat removal rate is accomplished without the requirement for external (manual or automatic) control. A thermohydraulic model is developed and applied to evaluate the proposed RHX with the existing conventional NDHX. Compared to the conventional design, the RHX has the potential to reduce heat removal by approximately 30 % at low temperatures and near-solidification conditions. In the case of the prototypical 20 MW reactor with a decay rate of 0.2 MW, RHX extends the freezing time of molten salt by three times, to approximately 18 h. During high-temperature events, such as the initial stages of a reactor scram, the RHX can improve decay heat extraction by 17 %. Other applications of molten salts with similar challenges can also be addressed using the proposed design approach.

Analysis of representative unprotected accidents for a sub-critical lead–bismuth eutectic reactor using IMPC-transient

For an accelerator driven sub-critical system (ADS), the transient analysis of the sub-critical reactor is very important. It is better to use the 3D neutron transport kinetic model rather than the 0D point reactor kinetic model due to the change of the neutron spatial distribution in some conditions. The code IMPC-transient has been developed to provide an effective system safety analysis tool for Chinese ADS project, in which3D neutron transport kinetics is available. A 10 MW ADS conceptual scheme, in which the target power is 2.5 MW and the reactor power is 7.5 MW, was proposed to test the applicability of IMPC-transient in sub-critical lead-bismuth eutectic (LBE) reactor. The steady state and three accident cases including unprotected loss of flow(ULOF), unprotected beam overpower(UBOP) and unprotected beam-trip(UBT) have been simulated, and the dynamic thermal-hydraulic response was focused on. The results show that IMPC-transient has the function of analyzing the ULOF, UBOP and UBT conditions for the sub-critical LBE reactor, and the 7.5 MW reactor conceptual scheme is feasible under steady state and has remarkable resistance against ULOF, UBOP and UBT.

New exposure room shielding incorporated with ferro boron concrete for neutron radiography imaging (NURI) facility at TRIGA PUSPATI Research Reactor

Simulation of the new exposure room shielding for Neutron Radiography Imaging (NURI) at 1 MW PUSPATI TRIGA MARK II (RTP) nuclear research reactor in Malaysian Nuclear Agency, Malaysia has been presented in this article. Previous exposure room shielding has to be disassembled due to various reasons such as radiation streaming problem and outdated since the exposure room was constructed in 1985. The NURI facility had been undergoing major upgrading for all their instrument including beam collimator, imaging system, sample positioning, instrumentation shielding, and exposure room shielding. This paper aims to simulate the radiation dose rate mapping for the new exposure room shielding for the NURI facility using the Monte Carlo simulation code of MCNPX. The exposure room shielding design consists of several modular blocks built from ordinary concrete and Ferro boron concrete. Ferro boron concrete is used to enhance the shielding performance of the facility. Results showed that exposure room shielding containing Ferro boron concrete manage to reduce the radiation by 1150 % at the critical area compared to ordinary concrete. The average radiation dose rate around the facility during operation at 1 MW reactor power ranging from 1 μSv/hr - 10 μSv/hr.

Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor

The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.

Preliminary assessment of the impact of extrapolation distance on the results of Time-Average calculations in CANDU analyses

The time-average calculations are typically used in CANDU full-core fuel management to capture the average behaviour of the core over a long period of time. This approach is necessary since CANDU reactors, which utilize natural uranium fuels, need to be refueled on a daily basis (or a few times a week) in order to keep the reactor critical. Solving the neutron diffusion equations for a time-average core is no different than for any snapshot of core configurations. The only difference is related to the burnup definition of the fuel bundles where instead of having a single value for each bundle, various other parameters are taken into account such as the fueling scheme (i.e., the number of bundles fueled during each refueling operation), the target exit irradiation or burnup, and the dwelling time (i.e., the duration spent by a bundle in one position in the core). Details regarding time-average calculations will not be discussed here. Instead, one of the components for a typical full-core physics calculation will be evaluated in more detail, namely the extrapolation distance which represents the distance from the edge of the core where the neutron flux goes to “zero”. The impacts of varying the extrapolation distance on the fueling rates and on the worth of the liquid zone controller are examined. The results of the study indicate that the total number of fuel bundles needed to support an operation of a CANDU 600 MW reactor for a year could vary by approximately ± 70 bundles and the worth of the liquid zone controller can vary by approximately 0.18 mk. Finally, it should be noted that the results presented herein are part of the initial assessments for developing a fuel-channel-dependent extrapolation distance for CANDU application.

ИССЛЕДОВАНИЯ ВЫСОКОТЕМПЕРАТУРНОЙ ЯДЕРНОЙ ЭНЕРГОТЕХНОЛОГИИ ДЛЯ ПРОИЗВОДСТВА ВОДОРОДА И ДРУГИХ ИННОВАЦИОННЫХ ПРИМЕНЕНИЙ

The presented results of neutron-physical, thermophysical, and technological studies have shown that it is possible in principle to provide the required parameters of a high-temperature (900-950 °C) reactor installation with a 600 MW reactor (heat) with a sodium coolant for hydrogen production and other innovative applications based on one of thermochemical cycles or high temperature electrolysis with a high coefficient of thermal use of electricity. The relative small size, type of coolant, and the choice of fissile material and structural materials make it possible to create a reactor with intrinsic properties that provide enhanced nuclear and radiation safety. The possibilities of using heatresistant radiation-resistant structural materials and using the proposed technology of hightemperature sodium coolant at a high hydrogen concentration are discussed. The features of the behavior of a high-temperature complex multicomponent heterogeneous system ″sodium coolant - impurities - structural (technological) materials - protective gas″ are analyzed, which are associated with an exponential temperature dependence of constants characterizing heat and mass transfer processes (diffusion, permeability, solubility, absorption rate, equilibrium gas pressures, etc.). It has been shown that the technology system for high-temperature sodium coolant differs significantly from systems at modern nuclear power plants. With an increase in the intensity of hydrogen sources coming from the third circuit to the second, compared with BN-600 NPPs by two to three orders of magnitude, the condition for its implementation is to increase the hydrogen concentration in sodium by two to three orders of magnitude compared to modern nuclear power plants in combination with the removal of hydrogen from sodium by evacuation through membranes from vanadium or niobium.

Electronic Simulation of Neutronic Transient Studies for PWR‐900 MW(e)

To study different functionalities of the PWR‐900 MW(e) reactor core, an electronic model of Simulation Program with Integrated Circuit Enfesied (PSPICE) for electronic circuit design was adopted to model the mathematical models. A classical breadboard electronically simulates the physical terminal characteristics of each network. The electronic simulator, parity simulator, for reactor core can simulate six‐delayed neutron groups, single prompt neutron, neutron source, fuel channel, the coolant, reactor poisons, temperature feedback, control rods, safety rods, and boron injection. Positive/negative step or sin reactivity functions are two case studies were analyzed. The obtained results were found to be in a good agreement with the available data of El‐Dabaa PWR 900 MW(e) basic simulator.

Iodine analysis of foodstuffs samples using epithermal instrumental neutron activation analysis

Iodine is an essential element that has an important role in the proper growth and thyroid hormone functioning of humans. Iodine deficiency was correlated with the health status of human-caused by malnutrition. The analytical method to analyze the Iodine content on foodstuffs is limited because Iodine concentration exists in a level of a trace. 127I has a ratio of I 0 to s relatively high, and it is possible to obtain 128I (t1/2 = 24.5 minutes) using nuclear reaction with an epithermal neutron. Samples were collected at Magelang region, and it consists of foodstuffs that consumed by the citizen. All sample has been dried using Freeze dryer at-90 °C and 0.03 bar for 24-48 hours. After drying, the sample then pounded to get the form of granules smooth, about 100-150 mesh. About 40 to 130 mg of sample was weighed accurately on a clean micro vial of Polyethylene. Irradiation has been carried out at l5 MW reactor power using neutron epithermal for 45 seconds at the rabbit facility of GA. Siwabessy reactor. After cooling 5 to 10 minutes, the irradiated sample then counted for 300 seconds, and the gamma-ray spectra obtained emitted have been analyzed by Hyperlab software. Several standard reference materials of NIST have been used as analytical quality control. The experimental result shows that the Iodine has the Boron ratio, RB, and the Advantage factor, Fadv, equal to 2.4 and 21, respectively. Trace element of Iodine has been determined quantitatively on nine spices, seven types of meat, and 32 vegetables. The Iodine concentration on spices have average 1.57 ± 0.23 mg/kg with range value 0.38 mg/kg to 3.4 mg/kg. Iodine on vegetable have average of 2.00 ± 0.43 mg/kg with range value of 0.38 mg/kg to 7.30 mg/kg. Meanwhile, its concentration on meat has range 0.71 mg/kg to 2.16 mg/kg, with an average of 1.44 ± 0.28 mg/kg, lower than that on vegetables. Epithermal INAA is a reliable analytical technique for Iodine determination quantitatively.

Confirmation of Nuclear Heating Rate for Installation of Cold Neutron Source at HANARO

In order to determine the capacity of the cold neutron source refrigerator of HANARO, the nuclear heating rate at CN vertical hole is measured by using the heat-flow calorimetric method and confirmed by the calculation. The heating rate measurement device of HANARO was composed of a calorimeter sensor, an air containing aluminum sleeve for fitting the sensor to the CN hole, aluminum weight and a lead wire assembly. The calorimeter sensor consists of a cylindrical Al sample and container, two thermocouples and the electric heater for the calibration of the calorimeter. The sample is separated by an air gap from the Al container surrounded by an air containing Al sleeve. After installation of the calorimeter at a measurement position of HANARO, the heat transfer inside the calorimeter was simulated by the electric heating for the sample. The nuclear heating rates at the CN hole were determined at three reactor powers of 1, 4 and 8 MW by using the calibration curve and the temperature measurements at each reactor power. The measured nuclear heating rate per unit mass of Al sample at 8 MW reactor power is 0.143 W/g and it is equivalent to the 0.494 W/g at 30 MW. The nuclear heating rate was calculated by using the MCNP code. The calculation model for the whole facility including the reactor core and the reflector tank were established. In the calculation procedure, the heat generations by various radiations were evaluated with considering the prompt, delayed and activation effects. The measured heating rate was reasonably well supported by the calculation using the cold neutron facility design code. It will be very useful for the moderator cell of cold neutron source of HANARO.

PERFORMANCE ANALYSIS OF RDE ENERGY CONVERSION SYSTEM IN VARIOUS REACTOR POWER CONDITION

Reaktor Daya Eksperimental (RDE) is an experimental power reactor based on High Temperature Gas-cooled Reactor (HTGR) technology with thermal power of 10 MW. As an experimental power reactor, RDE is designed for electricity generation and provides thermal energy for experimental purposes. RDE energy conversion system is designed with cogeneration configuration in the Rankine cycle. To ensure the effectiveness of its cogeneration, the outlet temperature of the RDE is set at 700°C and steam generator outlet temperature is around 530°C. Analysis of the performance of the energy conversion system in various power levels is needed to determine the RDE operating conditions. This research is aimed to study the performance characteristics of RDE energy conversion systems in various reactor power conditions. The analysis was carried out by simulating thermodynamic parameter calculations on the RDE energy conversion system and the overall cooling system using the ChemCad program package. The simulation is carried out by increasing the reactor power from 0 MW to 10 MW at constant pressure and constant mass flow rate. The simulation results show that the steam fraction at the steam generator outlet increases starting from 3 MW reactor power and reaches saturated steam after the thermal power level of 7.5 MW. From the results, it can be concluded that with constant mass flow rate and operating pressure, optimal turbine power is obtained after the reactor thermal power reached 7.5 MW.Keywords: RDE, Energy Conversion System, Performance, Reactor Power, ChemCad

CFD simulation and analysis of reactor integral hydraulic tests

Scaled-down reactor models are widely used to investigate the reactor integral hydraulic characteristics, due to the feasibility and economy. However, no definite conclusion can be drawn from previous researches for the effects of different scaling methods on simulation results. In this paper, in order to analyze the effects of simulation core design parameters and flow flux on core inlet flow distribution, a CFD simulation is conducted based on the hydraulic tests of the 1/4 scale model of 600 MW reactor in Qinshan phase II. The effects of different scaling velocities, core resistance distributions and lateral resistance on flow distribution are investigated. Calculation results reveal that the uniformity of flow distribution increases with test velocity. Ignoring the resistance of fuel assembly would reduce the uniformity of flow distribution. The flow distributions under conditions of uniformly and segmented distributed core resistance are basically the same, indicating that the detailed simulation of axial pressure drop distribution in assemblies is unnecessary for the design of hydraulic simulator of assembly. The lateral resistance could significantly affect the flow distribution and excessive lateral resistance would reduce the uniformity of distribution.

Aplicação da lâmpada de descarga de mercúrio sem eletrodo para degradação do paracetamol

Lâmpada de descarga de mercúrio sem eletrodo (Hg-EDL) foi aplicada em estudos de degradação fotolítica e fotocatalítica para avaliar o potencial de degradação de uma solução de Paracetamol 10 mg L-1. A fotodegradação (fotólise e fotocatálise) foi conduzida em reator UV/MW, com potência microondas fixa (200W, pH ̴ 7) e diferentes tempos de irradiação (0,083 a 2,0 min.). Após irradiadas, as amostras foram analisadas através de espectrofotometria UV/Vis para quantificação do fármaco. Houve a remoção de até 70,1% no ensaio fotolítico com o tempo de 2 min., enquanto para o processo fotocatalítico (aplicando suspensão de 1 g L-1 dos semicondutores) os resultados de remoção foram da ordem de 60%. A cinética de primeira ordem foi aplicada, sendo determinada uma constante k = 0,602 min-1 e r2 = 0,993, evidenciando um ajuste adequado a cinética de ordem 1. Assim, o reator UV/MW demonstra elevada eficiência no processo de degradação do paracetamol, uma vez que elevada taxa de remoção e valor elevado de constante cinética de degradação foram obtidas para o respectivo sistema. O foto reator avaliado pode ser explorado em ensaios de degradação com outros compostos, possibilitando a avaliação de sua eficiência frente a degradação de uma maior variedade de contaminantes emergentes.

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

Assessment of N-16 activity concentration in Bangladesh Atomic Energy Commission TRIGA Research Reactor

Abstract An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was 7 . 40 × 10 5 Bq / cm 3 and at the bay region was 3 . 39 × 10 − 5 Bq / cm 3 . At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was 4 . 14 × 10 − 1 Bq / cm 3 .

Anti-neutrino flux in a research reactor for non-proliferation application

Abstract Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 × 1017 ( ν ¯ e ${\bar \nu _e}$ /sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of 239Pu to the fission using an antineutrino detector is not viable in the TRR.

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Abstract In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1X-A1, U3O8-Al, and U3Si2-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1X-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U3Si2-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.