Neutronic Parameters Research Articles

The effect of enriched gadolinia and its concentrations on the neutronic parameters of AP-1000 fuel assembly

This paper investigates methods of effective use of gadolinium-bearing rods as burnable poisons in AP-1000 fuel assembly. To do this, a sensitivity analysis was performed on the effect of using the enriched Gd2O3 and various concentrations of Gd2O3 fuel pins. Different neutronic parameters of AP-1000 FA (that is widely used in pressurized water reactors) such as Kinf, peak pin power, reactivity swing, and the residual binding at different burnups are analyzed. The main goal of this investigation is to effectively use burnable poisons to maintain the reactivity swing as low as possible to decrease the frequent moving of the control rods. DRAGON code is used to develop subassembly designs. The cases with enriched Gd157 have burned out near 11 GWd/MTU and for Gd155 cases, it was near 17 GWd/MTU. The reactivity change of cases with enriched Gd155 is smoother than the cases with enriched Gd157. Our reactor physics study arrives at the flattest reactivity curve and the highest burnout time with the highest value of Gd155 enrichment (100 w/o). This design gives the lowest reactivity swing as well as a better pin power peaking factor. The new suggested design of enriched gadolinia bearing rods performed excellently in less reactivity swing, power peaking factor, and reactivity binding.

Searching for the most optimum burnable absorbers (BAs) for AP-1000 from the neutronic, thermal-hydraulic, and solid mechanics points of view

Because of the importance of absorber materials in regulating the reactivity of AP-1000, several materials with varied designs have been investigated to find the most optimum absorber materials from the neutronic, thermal-hydraulic, and solid mechanics perspectives. In this work, ZrB2, Er2O3, (SiO2-B2O3) and (Al2O3-B4C) were investigated as burnable absorbers (BAs) and H3BO3, Ag–In–Cd, and SS-304 as non-burnable absorbers. For the investigated BAs, the main neutronic parameters such as the fuel burn-up, axial thermal neutron flux distribution and the pin by pin distribution have been simulated. For the non-burnable absorbers, the control rod worth (CRW) and the boron concentration coefficient have been calculated. In the thermal-hydraulic analysis, the temperature distribution of the fuel, clad, coolant and the minimum departure from nucleate boiling ratio (MDNBR) have been studied. The BAs’ effects on solid mechanics analysis have emerged on both the von-Mises stress and the maximum fuel outer surface displacement.

Neutronic and thermal analysis of fuel rod containing light water cell in a fusion-fission hybrid reactor

In this study, the most suitable diameter and pitch of fuel rods, according to neutron parameters, were determined by MCNPX code. Due to the fact that by decreasing the temperature in the fuel rod, neutron production increases in the fission process, a light water cell is created in the center of the fuel rod. The temperature reduction in the center of the fuel rod increases effective multiplication factor keff and reactor's ultimate energy Efis. Temperature changes in the light water cell and in the fuel rod are analyzed theoretically and simulated by Ansys software. Finally, the infinite multiplication factor k∞, keff, the sum of the produced neutrons in the reactor, the total released fission energy per each fusion neutron and Efis examined in the selected design at temperatures of 293 K, 600 K and 900 K for fuel and temperatures of 293 K and 600 K for light water cell.

Comparison of radiation shielding and elastic properties of germinate tellurite glasses with the addition of Ga2O3

The current research elucidates the nuclear shielding capacity of germinate tellurite glasses: 41.7GeO2–41.7TeO2–16.6Ga2O3, 37.5GeO2–62.5TeO2, 10.4GeO2–72.9TeO2–16.7Ga2O3 and 12.5GeO2–87.5TeO2. Gamma-ray photon, fast neutron and electron shielding parameters of the present glassy materials were evaluated and studied via the Geant4 Monte Carlo, Phy-X/PSD software, ESTAR and analytic computations. In addition, Makishima–Mackenzie’s theory was applied to assess the elastic properties of the studied tellurite glass system containing Ga2O3 and/or GeO2. The effective atomic number of the glasses varies from 19.14 to 44.08 for 41.7GeO2–41.7TeO2–16.6Ga2O3, 20.63–48.02 for 37.5GeO2–62.5TeO2, 21.15–48.15 for 10.4GeO2–72.9TeO2–16.7Ga2O3 and 22.42–50.29 for 12.5GeO2–87.5TeO2. The obtained fast neutron removal cross sections of the glasses were 0.0991, 0.0966, 0.1024 and 0.1021 cm−1, respectively, for 41.7GeO2–41.7TeO2–16.6Ga2O3, 37.5GeO2–62.5TeO2, 10.4GeO2–72.9TeO2–16.7Ga2O3 and 12.5GeO2–87.5TeO2. Also, an equilibrium is reached between total stopping power (TSP) due to radiation and collision for electrons at energy T = 1.0 MeV where the TSP was minimum in the investigated glasses. Computed Young’s modulus for 37.5GeO2–62.5TeO2 was the lowest with a value of 0.218 GPa while the other three glass samples have almost equal value of 0.226 GPa. The present glasses’ shielding ability outclassed some conventional shields, hence have potential for radiation safety/shielding purposes in nuclear facilities.

Conceptual design of an innovative I&XC fuel assembly for a SMR based on neutronic/thermal-hydraulic calculations at the BOC

Abstract Power upgrade in nuclear reactors has been identified as one of the least costly options. This article focuses on how to further increase the thermal power and the possibility of using internally and externally cooled (I&EC) fuels instead of the solid fuels in the core of a Small Modular Reactor (SMR). In hence, The NuScale is chosen as the reference SMR. The core of NuScale is designed based on the use of a new 12 × 12 I&EC fuel assembly. This study is conducted throughout neutronic/thermal-hydraulic analysis. And many essential neutronic and thermal-hydraulic parameters such as variations of effective multiplication factor as a function of the pitch, neutron flux, power peaking factors, DNBRs and maximum temperature of the fuel were obtained. As one of the most important results of the analysis, I&EC fuel shows a sufficient margin available on DNBR and fuel pellet temperature compared with cylindrical solid fuel. The margin amount seems accommodating a 183% power-uprate seems viable. Also, the fuels axial temperature at different power levels were analyzed, and it was found that the proposed fuel at high power levels has a low peak temperature.

DETERMINATION OF THE NEUTRON CONTAMINATION DURING BRAIN RADIOTHERAPY USING A MODERATED-BORON TRIFLUORIDE DETECTOR AND THE MCNP MONTE CARLO CODE.

This study aimed to determine the neutron dose equivalent to the thyroid gland and eye lens in brain tumor radiation therapy with 15- and 18-MV three-dimensional conformal methods (3D-CRT). A Monte Carlo simulation was performed using the Monte Carlo N-particle transport code to calculate neutron fluence and ambient dose equivalent (H*(10)). Afterward, these parameters were measured using a model NRD roentgen equivalent in man (REM) neutron detector (Thermo Electron Corporation, USA) equipped with Eberline's ASP-2e rate meter. Finally, the organ neutron dose equivalent was obtained by applying depth corrections to the measured ambient dose equivalent at the distance of the organ center from the central beam axis. The ratio of the out-of-field photon dose equivalent, measured previously, to the neutron dose equivalent in the eye lens was high due to its proximity to the radiation field. In contrast, this ratio remained unexpectedly high in the thyroid gland that is far from the central beam axis (about 15cm). The calculated neutron parameters agreed with the measurements. The present study findings indicate that external field photon dose is the main source of thyroid gland biological effects in radiotherapy of brain tumors. In addition, it is appropriate to apply the model NRD REM neutron detector for measuring neutron contamination from high-energy linear accelerators inside and outside the treatment field.

The effect of adding impurities to the high density fuel matrix on the reactivity of the RSG-GAS core

The safety calculation of the RSG-GAS reactor core needs to be done if new high-density fuel is used in the RSG-GAS core. The use of new high-density fuels is very much needed in the RSG-GAS core so that it is necessary to calculate its safety. When using high density fuel there will be a “bottle neck” effect where there is a problem with the fuel and cladding interaction. According to previous researchers this can be overcome by adding metal elements Si, Ni or Ti to fuel. However, with the addition of metal elements Si, Ni or Ti to the fuel, it needs to be analyzed in the RSG-GAS core. The analysis was carried out calculations using a computer program by simulating the Si metal used 1%, 2%, 3% 4% and 5% in the fuel. The calculation of the additional metal element in the high-density fuel effect on the reactivity of the RSG-GAS core was carried out with the WIMSD-5B and Batan-FUEL programs. The WIMSD-5B program is a computer program consisting of several modules including the transport module which is used for the calculation of the generation of macroscopic constants for the 19.75% enriched uranium silicide fuel and the density of 4.8 gU/cc as a function of degree of burn. Core calculations were performed using a neutron diffusion program, namely Batan-FUEL. By calculating the core criticality at operating temperature, the core neutronic parameter values are obtained. The results of the analysis show that the amount of Si 5% in the high density fuel matrix can solve the bottle neck problem and the neutronic parameters are still within the safety limit. This value will determine how much the value of the fuel burn up and control rods are in accordance with the acceptance criteria so that the safety analysis can be carried out on the RSG-GAS core using high density. From the analysis, it was found that the addition of the element Si in the high-density fuel matrix did not find any significant changes in core reactivity and neutronic parameter.

Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

Design of the CAREM nuclear reactor core with dual cooled annular fuel and optimizing the thermal-hydraulic, natural circulation, and neutronics parameters

The small modular nuclear reactor (SMR) with annular fuel which is cooled internally and externally has the potential to maintain high power density with a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant from both the outer and inner surfaces.In this paper, design of the Argentina small modular reactor (CAREM) core with hexagonal assemblies using dual-cooled annular fuels with internal and external cooling is presented. Neutronics and thermal-hydraulic parameters are evaluated and analyzed by changing internal fuel radius. For this purpose, at the first, the dual-cooled annular fuel under clean and cold conditions is modeled and the effective multiplication factor has been calculated for different inner clad diameter. Indeed, the internal and external radius have been changed in such a way to maintain the reactor under moderated.Then, these annular fuels under full power conditions are modeled and power peaking factor has been calculated. Finally, natural circulation parameters are performed for a simulated fuel rod in the hot channel using computational fluid dynamics (CFD) simulation codes. These results are compared with the conventional CAREM reactor. One of the most prominent advantages of the annular fuels is the ability to make softer neutrons in which reduce maximum power peaking factor and improves fuel management and safety parameters in the reactor core.For data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses and output consists of fuel rods’ pitch, thermal-hydraulic and neutronic parameters. Finally, the optimal geometry of fuels is determined using the neural network by implementing the genetic algorithms.Indeed, developed Artificial Neural Network (ANN) utilizing the obtained data, predicts the thermal-hydraulic and neutronic parameters of the CAREM reactor core with dual-cooled annular fuels. Presented optimization algorithm, which has a significant ability to attain the best solutions, also determines the optimal values of natural circulation parameters (Vmax/Vavg, Vout-Vin, and pressure drops), heat transfer coefficients, MDNBR, RPPF, and excess reactivity of the reactor. Also, the designed artificial neural network and genetic algorithm has been validated using neutronic and thermal hydraulic calculations. Results indicate that by using annular fuels, better reactor safety and thermal efficiency are achieved.

Approach for the adaptations of a nuclear reactor model towards more flexibility in a context of high insertion of renewable energies

The massive penetration of renewable energy sources (RES) that are variable and not “dispatchable”, may weaken the power system supply-demand balance. Nuclear power plants (NPP) contribute in part to this daily and seasonal balance thanks to the “load-following” mode in France for example, but there are still limits to their use. These limits prevent a nuclear power modulation as efficient and quickly as the conventional thermal power plants. The need in terms of power ramps for nuclear in a constrained power system has been quantified in previous studies. Nuclear may compensate for the removal of thermal power plants, in order to fulfill energetic strategies of CO2 reduction. The possibility that nuclear reactors can achieve power ramps of significant values (>5%Pn/min) is put forward and could make possible to replace the services currently provided by thermal power plants. The objective of the study is then to use these power system requirements as the main input parameter for the modelling of a current simplified nuclear reactor capable of responding to frequency control within a specific hypothesis framework. In this paper, a French 1300 MW pressurized water reactor is modelled. Parametric studies are carried out in order to reveal technical and technological constraints when increasing electric power ramp. The study explores ways of design, which may influence reactor flexibility, such as the neutron parameter, Doppler coefficient, or the thermohydraulic parameter, delay in the primary loop.

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup.

Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors

In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

Neutronic feasibility study of ([formula omitted] fuel in a modelled TRIGA research reactor

This paper presents a comparative analysis when transitioning from fuel meat based on uranium zirconium hydride (U-ZrH1.65) to one based on thorium-uranium-233 zirconium hydride ((Th-233U)-ZrH1.65) as an alternative option for TRIGA reactors. The aim of this study was to investigate the neutronic characteristics of (Th-233U)-ZrH1.65 fuels with two different Th-233U weight contents (i.e., 8.5 wt% and 12 wt%) and various U-233 enrichment levels. Using MCNP6.2 computer code based on the Monte Carlo method, a three-dimensional (3D) hypothetical TRIGA model was developed to investigate various neutronic parameters, such as the effects of U-233 enrichment and different (Th-233U) weight contents on the effective multiplication factor, burnup behavior, spent fuel composition, neutron flux, and fission product poison concentration. Since fuel temperature coefficients of reactivity and effective delayed neutron fractions play important roles in the reactor’s kinetics, these were also investigated, and the results were compared to the model with U-ZrH1.65 fuel. The obtained results revealed that the presence of Th-232 in the fuels did not drastically reduce the fuel temperature coefficients due to the zirconium hydride within the fuels providing most of the feedback in a TRIGA reactor. In contrast, the presence of U-233 in the (Th-233U)-ZrH1.65 fuels did reduce the effective delayed neutron fraction. Despite this, the lower transuranic generation in cores fueled by (Th-233U)-ZrH1.65 make this fuel advantageous in terms of achieving longer cycle lengths and less-hazardous nuclear waste for disposal.

Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm

Abstract The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values.

Fissile utilization of uranium, thorium, and plutonium fuels for a 60 MWt block‐type high‐temperature gas‐cooled reactor

High-temperature gas reactor (HTGR) is one of the Generation IV nuclear reactors currently undergoing rapid development in the world. The utilization of fissile material becomes an important parameter in HTGR technology because of the use of layered particle-type fuel that makes it challenging to reprocess. This research aims to compare and optimize the fissile utilization of HTGR 60 MWt using UO2, (Th-U)O2, and (Pu-U)O2. The utilization of fissile material on HTGR is being studied in this research by using the SRAC (Standard Reactor Analysis Code) version 2006 and the Japanese Evaluated Nuclear Data Library (JENDL) version 4.0. Parametric surveys are performed for important neutronic parameters, including analyzing the effects of fuel enrichment and the coated fuel particle (CFP) amount. Fissile materials are analyzed on a 1% to 20% enrichment level, and the CFP number is varied from 4470 to 26 830. The neutronic parameters taken into account are k-inf, conversion ratio (CR), the density of fissile and fertile material, and neutron energy spectrum. The results suggest that 11% of fuel enrichment and CFP number between 8950 and 11 180 are optimal for 60MWt HTGR. The highest fuel utilization level at the end of the reactor period was achieved by Pu-based fuel, with 81.6% of utilized fissile material.

Analyzing the Neutron Parameters in the Accelerator Driven Subcritical Reactor Using the Mixture of Molten Pb-Bi as Both Target and Coolant

In this paper, the Accelerator Driven Subcritical Reactor (ADSR) was simulated based on the structure of the TRIGA-Mark II reactor by the MCNPX program. The proton beam interacts on the Pb-Bi molten target with various energy levels from 0.5 GeV to 2.0 GeV. The important neutron parameters to evaluate the operability of ADSR were calculated as: the neutron yields according to various thicknesses of the target and according to the energy of the incident proton beam; the effective neutron multiplication factor for various fuel mixtures, along with its stability for some fuel mixtures; the axial and radial distributions of the neutron flux along with the height and radius of the core. The obtained results had shown a good agreement in using Pb-Bi molten as the interaction target and coolant for ADSR.

Development and application of TaSNAM 2.0 for advanced pressurized water reactor

The Nuclear Thermal-hydraulic Laboratory (NutheL) at Xi’an Jiaotong University developed a Temporal and Spatial Neutronics Analysis Module (TaSNAM) based on the User Defined Function (UDF) and User Define Scalar (UDS) in Fluent. The neutron diffusion equations were solved based on the Finite volume method (FVM). In this paper, the TaSNAM is extended the original function to solve 3D problems and obtain neutron parameters based on the neutron parameters library, and improved to TaSNAM 2.0. The 3D-IAEA benchmark problem was selected and simulated to validate. The calculated results show good agreement with the reference values and the difference from the reference value of keff is 40pcm. A three-dimensional steady-state simulation of the whole core in AP1000 was completed employing the TaSNAM 2.0. The calculated results of important parameters agree well with the design values and the detailed three dimensional distributions of neutron flux and power were achieved.

Neutronic modeling and calculation of the Nuclear Heating Reactor NHR-5 by the deterministic codes DRAGON5 & DONJON5

The main objective of this study is to assess the neutronic modeling and calculations of the nuclear heating reactor NHR-5. To this end, the neutronic parameters of NHR-5 reactor underwent a comparative and validation protocols/methods through the analysis of the finite multiplication factor keff and some neutronic parameters including D,νΣf,Σa as well as the power and flux distributions, using the evaluated nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0 based on 172 energy groups.In this study, the collision probability approach is used to simulate and calculate the neutronic parameters in NHR5 reactor using the deterministic core diffusion code DONJON5 and the lattice transport code DRAGON5. The group constants of the reactor components were produced by DRAGON5. The DONJON5 code was then used to calculate the effective multiplication factor, excess reactivity, and power and flux distributions by introducing these group constants. The results are compared to those of the experiment to validate the calculation scheme. The results of the simulations reveal that the computed neutronic parameters consistently produce reasonable and consistent results of flux and power distributions, excess reactivity and all other parameters when compared to the experiment values. Therefore, the reactor models of NHR5 developed by DRAGON and DONJON codes were good in predicting the effective multiplication factor as well as the studied neutronic parameters. In this study, we being able to show the potential validation of the reactor physics lattice transport code DRAGON5 and the core diffusion code DONJON5, as well as the nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0.

Core Configuration Analysis for Modular Gas-cooled Fast Reactor (GFR) using OpenMC

The core configuration analysis of modular Gas-cooled Fast Reactor (GFR) has been done to understand GFR performance. The modular GFR used a fast neutron spectrum and high-temperature helium gas, providing higher thermal efficiency than the other generation IV reactor candidates. In this paper, the variation of core configuration and dimension for core design has been applied in radial, axial, and radial-axial directions. The Monte Carlo method, named OpenMC code, has been used for the criticality and isotope evaluation of design core GFR. The OpenMC code provides the probabilistic solution to solve the neutron transport equation in a 3D model and non-homogenous physical volumes using Evaluated Nuclear Data File (ENDF/B-VII.b5) and continuous energy spectrum. The neutronics parameters characterized are the value of keff, fission rate and neutron flux distribution, and fissile material evolution to know of GFR core design’s performance. The analysis showed that the core configuration in radial direction gives a good understanding of the feasibility of GFR core design.

Examine the possibility of increasing the plutonium incineration rate in the current operating pressurized water reactor

This work investigates the possibility of increasing the incineration rate of reactor-grade plutonium (rgPu) in the pressurized water reactor (PWR). In an attempt to reduce the production of plutonium isotopes from the conventional fuel and increase the rate of burning the accumulated plutonium, thorium-232 is suggested to be used as a fertile material in the standard fuel assembly and a suggested Blanket-seed assembly. The neutronic properties of Th-rgPu in both the standard fuel assembly and the BS assembly were analyzed and compared with the standard fuel to distinguish which of them is more effective in burning the rgPu. The performance of Th-rgPu in the standard fuel assembly and the BS assembly has been assessed by tracking the infinity multiplication factor (Kinf), the amount of the rgPu incinerated, reactor safety parameters, the most important fission products, the radioactivity of the spent fuel and the radial power distribution through the assembly. The neutronic analysis confirmed the viability of BS assembly in burning a large quantity of rgPu, maintaining the neutronic safety parameters at acceptable values.