Neutronic Characteristics Research Articles

Comparative analysis between homogeneous and heterogeneous models of gas cooled fast reactor core (GFR-2400)

Abstract The main objective of the presented work is to investigate the effectiveness of homogenization of the GFR-2400 core in simulating its neutronic characteristics. To explore and evaluate the neutronic behavior of the large scale Gas cooled Fast Reactor GFR-2400 core, two computer models (homogeneous and heterogeneous) were designed using the MCNPX code. The designed heterogeneous model has been validated by comparing its results with a previously published paper. The results of both models were compared with each other to study the effect of fuel homogeneity on the radial flux and power distribution through the core. As part of a safety analysis of the reactor core for both designs, the reactivity worth of control rods, Effective delayed neutron fraction (ß eff), Doppler constant and Depressurization effect have been analyzed and compared. The variations of the effective multiplication factor (k eff), the minor actinides concentration and the most important fission products as a function of operation time have been investigated for the presented designs.

Ordered Particle Packing in Dense TRISO/SiC Fuel Elements and Preliminary Assessment of Neutronic and Thermomechanical Characteristics

Detailed analysis of the particle distribution in Transformational Challenge Reactor fuel elements indicates that particle packing is not random; instead, it follows a relatively ordered structure near fuel element surfaces. Discrete particle neutronic simulations indicate that the core reactivity is not impacted when assuming homogenization of particles with the silicon carbide matrix. However, the neutronic power distribution resulting from the ordered packing structure indicates that the highest-power particles reside at the top and bottom of the fuel elements and nearest the YH1.85 moderator rods. The power distribution results were applied to thermomechanical simulations using mesh-based power distributions. Previous results indicated high stress at the bottom of the fuel element, where packing is most ordered. To reduce this stress concentration, additively manufactured protrusions were added to the bottom of a test fuel element to disrupt dense particle packing. These protrusions reduced the overall power peaking, but the thermomechanical simulations did not indicate a significant change in the fuel element’s maximum stress or failure probability.

Method research and engineering validation of the improved homogenization for the heavy reflector in VVER

Compared with the traditional Cartesian PWR design, the heavy reflector, composed of larger amount of iron and less moderator, is an important feature of the VVER reactor. This structural feature would result in two different neutronics characteristics: first, less moderator would harden the neutron spectrum and make more adjoint reflector-node correlated; second, larger amount of iron within harden neutron spectrum may lead to notable resonance phenomenon for the structure–material isotopes. Therefore, the above neutronics characteristic would introduce corresponding challenges to the homogenization method for the heavy reflectors. In order to provide more accurate few-group homogenized parameters, an improved homogenization method has been proposed. As method verification, a two-dimensional VVER benchmark has been simulated, with the reference values provided by the Monte-Carlo code. Moreover, the improved homogenization method has been applied to the real VVER-1000 reactor operated in China, validating the simulation values of the key parameters with corresponding measurement values.

Analysis of Cosmogenic Neutron Characteristics and the Pulses Counting Rate Using ASD, LSD, and LVD Scintillation Detectors

Analysis of Cosmogenic Neutron Characteristics and the Pulses Counting Rate Using ASD, LSD, and LVD Scintillation Detectors

Fusion-fission hybrid reactor facility: neutronic research

The authors investigate the neutronic characteristics of the operating mode of a hybrid nuclear-thermonuclear reactor. The facility under study consists of a modified core of a high-temperature gas-cooled thorium reactor and an extended plasma neutron source penetrating the near-axial region of the core. The proposed facility has a generated power that is convenient for the regional level (60–100 MW), acceptable geometric dimensions and a low level of radioactive waste. The paper demonstrates optimization neutronic studies, the purpose of which is to level the resulting offsets of the radial energy release field, which are formed within the fuel part of the blanket during long-term operation and due to the pulsed operation of the plasma D-T neutron source. The calculations were performed using both previously developed models and the SERPENT 2.1.31 precision program code based on the Monte Carlo method. In the simulation, we used pointwise evaluated nuclear data converted from the ENDF-B/VII.1 library, as well as additional data for neutron scattering in graphite from ENDF-B/VII.0, based on the S (α, β) formalism.

Experimental research and analysis of VVER fuel composition with erbium oxide absorbers

NRC Kurchatov Institute's critical facility performed a series of experiments to study neutron characteristics of water-uranium compositions that contains fuel pins with uranium enriched up to 6.5 wt% of U-235 and erbium oxide (Er2O3) absorbers. Sizes, shapes, and compositions were defined to model uranium-erbium fuel considered for VVER extended fuel cycles. We believe the experimental results are valid for verifying and validating the computer codes employed to research and design advanced fuel cycles. The paper describes the main implemented fuel compositions and presents experimental results for criticality parameters and neutron flux distributions.The Monte Carlo code MCU-PD and the spectral code TVS-M were applied for the computational analysis. Both of these codes are used for VVER fuel cycles design. Results of the calculations of the primary neutron characteristics are compared to the experimental values.

On development of in-core fuel system for PWR reactors: Part I generation of macroscopic cross sections using SCALE 6.0 for use in nodal calculation

This work presents the description of the first part of a methodology applied to perform In-Core Fuel Management (ICFM) in Pressurized Water Reactor (PWR). The ICFM of a PWR reactor consists on defining the best charging or recharging pattern of fuel assemblies inside a reactor for an operational cycle. This means, finding a suitable arrangement of fuel assemblies that optimizes the performance of the reactor, which complies with all safety criteria. Genetic algorithms (GAs) are used to select the arrangements that interact with the reactor physics simulation code, holding the neutron characteristics of each fuel assembly. Therefore, a reliable and fast code was developed accordingly. The consolidated technique of coarse mesh node code that numerically solves the multigroup diffusion equation for two groups of energy, fast and thermal neutrons, in two dimensions was selected. In this type of code, it is essential that each fuel assembly is homogenized and characterized by its macroscopic cross sections, for each reactor’s burnup condition. The cross sections are generated with the support of SCALE 6.0, computational platform developed by the Reactor and Nuclear Systems Division (RNSD), from the Oak Ridge National Laboratory (ORNL). The completeness of the qualification and validation of the results obtained from the homogenization of the fuel assembly by the SCALE was performed comparing the results with actual data of a benchmark reactor. The fully documented Almaraz Nuclear Power Plant provided by the International Atomic Energy Agency (IAEA)-TECDOC-815, has been used as benchmark with successful results.

Conceptual design and thermal-hydraulic analysis of a megawatt-level lead-bismuth cooling reactor for deep-sea exploration

A new conceptual design of a mega-power nuclear reactor (MPNR) for applications in deep-sea exploration was developed. A solid core cooled by the natural circulation of a lead-bismuth eutectic (LBE) reactor was designed with a thermal power of 1 MW. It consists of 126 coolant channels, which are holes surrounded by highly enriched fuel of UO2 dispersed in titanium-zirconium-molybdenum (TZM) alloy. A control rod at the center of the fuel and six control drums around the block are made of boron carbide to control the reactivity of the reactor. The core and heat exchangers are arranged horizontally to reduce the volume of the entire system. Compared with traditional deep-submersible reactors used at sea, the MPNR is highly compact, has a high-power output and simple structure, and ensures safety under ocean conditions. The neutronic and thermal-hydraulic design characteristics of the MPNR were calculated and assessed. The temperature reactivity and void fraction reactivity coefficients of the MPNR are negative values to ensure that the reactor has negative feedback characteristics. The thermal-hydraulic analysis of the MPNR design was carried out using the computational fluid dynamics program STAR-CCM+, in which field functions were added to the program to realize the inclined, rolling, and heaving motions of the reactor. The numerical simulations showed that under stationary conditions, the MPNR could establish natural circulation with a mass flow rate of 93.32 kg/s. At an inclination angle of ±30°, the natural circulation mass flow rates are 88.6 kg/s and 87.5 kg/s, respectively. Under rolling and heaving conditions, the mass flow rate fluctuated periodically, and the maximum temperature of the core changed within 10 K.

Pulse characteristics of CLYC and piled-up neutron–gamma discrimination using a convolutional neural network

A method of describing the characteristics of Cs2 LiYCl 6 scintillator pulses is proposed by fitting all individual pulses with multiple exponential functions. Simulated pulses under very high counting rates (e.g., 2–40 Million counts per second) are generated by sampling from fitted parameter distributions, which include the non piled-up neutron (n) and gamma (g) pulses as well as piled-up gamma + neutron (g + n), neutron + gamma (n + g), neutron + neutron (n + n), and gamma + gamma (g + g) pulses. A fully connected neural network (FCNN) model can discriminate n and g signals clearly and demonstrates a discrimination performance similar to the traditional charge comparison method. A model trained with experimental data demonstrated very good prediction accuracy on simulated data, and a model trained with simulated data demonstrated good prediction accuracy on experimental data as well, indicating all characteristic of CLYC had been simulated. A convolutional neural network (CNN) model could discriminate between the six pulse types more effectively than a FCNN model, producing a discrimination error lower than 0.22% for n, g, n + g, and n + n pulses, and an error lower than 3.61% for g + g and g + n pulses. These results indicate that a CNN model is very suitable for complicated n–g discrimination under a piled-up condition.

Synthesis and characterization of vanadium(V) oxide reinforced calcium-borate glasses: Experimental assessments on Al2O3/BaO2/ZnO contributions

This study aimed to synthesize and characterize novel calcium borate glasses based on nominal compositions of 55B2O3-35CaO-(10–x)Al2O3-xV2O5, 55B2O3-35CaO-(10–x)BaO2-xV2O5 and 55B2O3-35CaO-(10–x)ZnO-xV2O5 (where x; 0.5 and 1 mole%). Experimental methods were used to determine the optical, structural, physical, and nuclear radiation shielding properties. Density values of the synthesized glasses were measured between 2.5681 g/cm3 and 3.0231 g/cm3. All glasses exhibit a transmittance characteristic start in the visible region. The transmittance starts at approximately 370 nm and this value shifts up to 420 nm. Moreover, neutron and gamma-ray shielding properties of all glasses were measured using experimental methods. Our findings showed that ZnO and V2O5 substitution was more effective than other substitution types (i.e., Al2O3/V2O5, BaO2/V2O5) in terms of optical band gap values. Our findings also showed that BCBV0.5 sample with the highest optical band gap has superior neutron and gamma-ray shielding properties. It can be concluded that substituting ZnO and V2O5 is a more effective method of increasing the crucial neutron and gamma-ray shielding characteristics of calcium borate glasses.

Measurement of the spatial-energy distribution of neutrons in the irradiation channel of the critical facility

One of the basic installations of the Republican State Enterprise “Institute of Nuclear Physics” of the Ministry of Energy of the Republic of Kazakhstan is a critical assembly, which is a zero-power reactor. Desalinated water and beryllium serve as moderators and neutrons reflectors. The energy spectrum of neutrons in the core is thermal. The main purpose and area of application is the modeling and study of the neutronic characteristics of the cores of water-moderated research reactors of various types.The paper presents the results of experimental measurements of the spatial-energy distribution of neutrons in the dry, central channel of the critical assembly. Measurements of the neutron flux were carried out using activation foils for three energy groups of neutrons: thermal, epithermal, and fast.The measured thermal neutrons flux in the irradiation channel is ~ 3·108 cm‒2s‒1, and fast neutrons flux (with energies above 0.7 MeV) is ~ 8·108 cm‒2s‒1. The fraction of thermal neutrons in the integral flux was 0.23%, and the fraction of fast neutrons was 0.62%. In the axial distribution of thermal and fast neutrons, the maximum value of the neutron flux is 50 mm below the midplane of the core.

Optical, structural, physical, and nuclear shielding properties, and albedo parameters of TeO2–BaO–B2O3–PbO–V2O5 glasses

In this study, glass systems were constructed according to the formula 34TeO2∙xBaO∙35B2O3∙(30-x)PbO∙1V2O5 (x=0, 5, 10, 15, 20, 30 mol%). The gamma, neutron, proton, and alpha shielding characteristics of the glasses were evaluated for nuclear security applications. The mass attenuation coefficient (MAC), half-value layer, mean free path, effective atomic number, and radiation protection efficiency (RPE) were experimentally measured in the energy range of 59.54–661.62 keV. In addition, the theoretical MAC values were determined. The exposure buildup factor, fast neutron removal cross section, mass stopping power, and projected range values for proton (H1) and alpha (He+2) particles were theoretically determined. Albedo parameters were measured for the prepared glasses at photon energy of 59.54 keV. The optical and structural properties were studied using X-ray diffraction, FT-IR Fourier transform infrared, ultraviolet–visible, and electron paramagnetic resonance spectroscopy. The glass samples comprised structural units such as TeO4, TeO3, BO3, and BO4. The spin Hamiltonian parameters, g∥, g⊥, A∥, A⊥, dipolar hyperfine coupling parameter, P, and Fermi contact parameter, κ, were calculated. The molecular orbital parameters, α2, β2, and γ2, were determined by optical absorption spectroscopy. The number of bonds per unit volume in the glasses ranged between 8.763 and 9.212×1028 m−3. The most reflective compound among the glasses investigated was BaO. Good agreement was found between the theoretical and experimental MAC values. TeBaBPbV1 glass was identified as a potential candidate for use in neutron, gamma, proton, and alpha shielding.

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.

Investigation of a new approach for regulating the reactivity and achieving economic feasibility using thorium in a blanket‐seed assembly of pressurized water reactors

The traditional methods for controlling reactor excess reactivity are expensive and have defects during reactor operation. By altering the fuel structure and fissile material distribution, this work presents a new technique for managing excess reactivity at the start of the fuel life. MCNPX code edition 2.7 was used to investigate the neutronic characteristics of the suggested blanket-seed assembly models and compare them with the reference PWR fuel assembly. Instead of U-238, Th-232 was suggested as a fertile nuclide. The impact of the proposed models on the infinite multiplication factor, fuel component concentrations, reactor-grade plutonium, minor actinides, radial flux of thermal neutrons, and the radial distribution of power has been analyzed. The safety parameters, including the control rods worth, Doppler effect, moderator temperature reactivity coefficients, and the effective delayed neutron fraction, have been studied to reach the most optimum fuel structure.

Neutronic Analysis of The SMART Modular Reactor Fuel Using SRAC 2006

Neutronic analysis of The SMART modular reactor fuel using SRAC 2006 has been carried out. Electrical energy is important today because the need is increasing along with the increase in human population, advanced technology and the economy. On the other hand, there are demands from the community for the clean, efficient and consistent energy. This is the reason why nuclear power plants are considered as one of the candidates for electrical energy suppliers in Indonesia in particular. This study evaluates a SMART reactor with Gadolinium as the burnable absorber material. The two kinds of fuel assembly were analyzed using the SRAC 2006 code system with the JENDL 4.0 as nuclear data library. This study aims to observe the neutronic characteristics of the fuel assembly designs according to the reference used. The results of the study show that of all types of fuel assemblies used can reach criticality at the beginning of the operating cycle and last up to 3 till 5 years when it finally reaches subcritical condition. Another parameter observed is the conversion ratio value, which from this study is in accordance with the characteristics of the conversion ratio for thermal reactors.

Calculation and Experimental Determination of the Neutronics Characteristics of the IRT-T Research Reactor

Calculation and Experimental Determination of the Neutronics Characteristics of the IRT-T Research Reactor

Core design and fuel behaviour of a small modular pressurised water reactor using (Th,U)O[formula omitted] fuel for commercial marine propulsion

There is an increasing interest in employing technologies to decarbonise the shipping sector; one option is the use of nuclear energy to power civilian ships. Previous work has investigated the use of UO2 fuel and its behaviour in a novel, soluble-boron-free, low power density PWR that offered a 15 year core life to power a container ship with a demand of 110 MWe. Here, building on that previously-presented UO2 core design, we investigate the analogous design of a (Th,U)O2 core to examine the neutronic and fuel performance benefits of using (Th,U)O2 fuel. The motivations for studying (Th,U)O2 fuel include understanding how (Th,U)O2 would behave in a novel low power density system and its behaviour as a fertile poison in a long-life core. Core design was developed using CASMO-SIMULATE and fuel performance was analysed using ENIGMA. In this new design, the introduction of (Th,U)O2 fuel exhibited similar neutronic characteristics (shutdown margins, xenon transient behaviour and reactivity coefficients) relative to the UO2 core. However, the fuel behaviour predicted by ENIGMA for (Th,U)O2 in this low power density core was found to be significantly impaired compared with using UO2 fuel. In particular, the introduction of uranium to ThO2 was found to degrade thermal conductivity and thorium-based fuels exhibited worse creep behaviour than UO2 fuels; both conclusions are supported by published experimental evidence. A factor behind the degraded fuel performance characteristics predicted by ENIGMA is the assumption that athermal phonon scattering by fission products has the same effect on thermal conductivity as for UO2. This assumption is necessary to address the limited burnup data for (Th,U)O2 fuels.

Study on the Concept of Organic-Cooled Microreactor

In order to study the concept of organic-cooled reactor, this paper firstly analyzes the important characteristics and key technical problems of the organic fluid used as the reactor coolant and moderator, as well as the main technical schemes for organic-cooled reactors. On this basis, this paper studies the neutronic characteristics of the 5 MW microreactor core. As shown in the study, under the same core layout conditions, the absolute value of the moderator temperature coefficient of the organic-cooled reactor is below that of the pressurized water reactor (PWR), and the power distribution is more flat. The study results from this paper can provide an important reference for the determination of the technology roadmap for multi-purpose microreactor power supplies and heating systems, which are rapidly developed in China.

Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis

The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle.

Steady State Calculations of the PWR MOX/UO2 Core with the Monte Carlo Code MCNP6

This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO2 (MOX: Mixed Oxide) Core Transient Benchmark with the modern MCNP6 Monte Carlo code based on the ENDF/B-VII.1 evaluated nuclear data library. The purpose was to verify an MCNP6 model proposed for calculations of a heterogeneous MOX/UO2 fuelled PWR core, which has different neutronic characteristics from the popular homogeneous ones loaded with the UO2 fuel due to its partial loading of the MOX fuel. The effective neutron multiplication factors, assembly power distributions, and control rod worths calculated using MCNP6 showed a reasonable agreement within 390 pcm, 6%, and 175 pcm, respectively, with the available benchmark data. The discrepancies between the MCNP6 results and the benchmark data were also discussed. Consequently, these results obtained with MCNP6 and ENDF/B-VII.1 can be considered as a new full-core heterogeneous transport solution to supplement for the available benchmark solutions at the steady-state conditions.