• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2564-2569
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• 2021

The neutronic analysis of Helium Cooled Ceramic Reflector (HCCR) breeding blankets has been performed using the 3D Monte Carlo code MCNPX and ENDF nuclear data library. This study aims to reduce ⁶Li percentage in the breeder zones as much as possible ensuring tritium self-sufficiency. This work is devoted to investigating the effect of ⁶Li percentage on the HCCR breeding blanket's neutronic parameters, such as neutron flux and spectrum, Tritium Breeding Ratio (TBR), nuclear power density, and energy multiplication factor. In the ceramic breeders at the saturated thickness, increasing the enrichment of ⁶Li reduces its share in the tritium production. Therefore, ceramic breeders typically use lower enriched Li from 30% to 60%. The investigation of neutronic analysis in the suggested geometry shows that using 60% ⁶Li in Li₂TiO₃ can yield acceptable TBR and energy deposition results, which would be economically feasible.

• Nuclear Engineering and Technology
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• 제50권7호
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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.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2126-2132
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• 2021

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m-3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.55-63
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• 2016

Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy ($E_e$) and source multiplication coefficient ($k_s$), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient ($k_{eff}$), net neutron multiplication (M), neutron yield ($Y_{n/e}$), energy constant gain ($G_0$), energy gain (G), importance of neutron source (${\varphi}^*$), axial and radial distributions of neutron flux, and power peaking factor ($P_{max}/P_{ave}$) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current ($I_e$) have been decreased in the highest case of $k_s$, but G and ${\varphi}^*$ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing $E_e$ from 100 MeV up to 1 GeV, $Y_{n/e}$ and G improved by 91.09% and 10.21%, and $I_e$ and $P_{acc}$ decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np-Pu assemblies on the periphery allows for a consistent $k_{eff}$ because the Np-Pu assemblies experience less burn-up.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3164-3170
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• 2021

The increasing interest that GEANT4 is gaining nowadays, because of its special capabilities, prompted us to address its reliability in neutronic calculation for the realistic and complex spallation target MEGAPIE of the Paul Scherrer Institute of Switzerland. In this paper we have specifically addressed the neutronic characterization of the active zone of this target. Three physical quantities are evaluated: neutron flux spectra and total neutron fluxes on target's z-axis, and the neutron yield as a function of the target's altitude and radius. Comparison of the obtained results with those of the MCNPX reference code and some experimental measurements have confirmed the impact of the geometrical and proton beam models on the neutron fluxes. It has also allowed to reveal the intrinsic influence of the code type. The resulting differences reach a factor of ~2 for the beam model and 4-18% for the other parameters cumulated. The analysis of the neutron yield has led us to conclude that: 1) Increasing the productivity of the MEGAPIE target cannot be achieved simply by increasing the thickness of the target, if the irradiation parameters are not modified. 2) The size of the spallation area needs to be redefined more precisely.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2678-2686
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• 2023

This work demonstrates the results of full-scale numerical experiments of a hybrid thorium-containing fuel plant operating in a state close to critical due to a controlled source of D-T neutrons. The proposed facility represented a level of generated power (~10-100 MW_t) in a small pilot. In this work, the simulation of the D-T neutron plasma source operation in conjunction with the facility blanket was performed. The fission of fuel nuclei and the formation of spatial-energy release were studied in this simulation, in pulsed and stationary modes of the facility operation. The optimization results of neutronic and fluid dynamics studies to level the emerging offsets of the radial energy formed in the volume of the facility multiplying part due to the pulsed operation of the D-T neutron plasma source were presented. The results will be useful in improving the power control-based subcriticality monitoring method in coupled systems of the "pulsed neutron source-subcritical fuel assembly" type.