Reactor Neutron Research Articles

Optimization of the route of the refueling machine to reduce the refueling time: Case study of the BN-800 reactor

Sodium-cooled fast reactors (SFR) are included in the list of fourth-generation nuclear energy systems, the so-called generation IV (GEN-IV). It is the only technology of GEN-IV that possesses the significant practical experience of design, construction, and operation of high-power reactors. Due to the high temperature of the coolant at the BN-600 and BN-800 reactor core outlets, the steam generator produces steam with higher enthalpy, and this significantly increases the efficiency of cogeneration at SFR power plants in comparison with pressured water reactor (PWR) or boiling water reactor (BWR) units. The primary nuclear fuel effectiveness rises due to cogeneration with higher efficiency of a sodium-cooled fast reactor nuclear power plant (NPP) in comparison with thermal reactors. In addition, it is necessary to reduce the reactor refueling outages to improve the nuclear power plant utilization factor.It is possible to reduce the duration of the reactor refueling by the route optimization of the refueling machine movement. The prevailing in the world thermal neutron reactors with water cooling have a refueling machine that points the certain fuel assembly using two coordinates. The fast neutron reactors use sodium as a coolant; thus, there is a problem with its violent reaction with water and air oxygen. Therefore, it is necessary to exclude the contact of sodium and surrounding air during the refueling. To achieve this, the system of refueling machine pointing is used, consisting of two or three eccentrically located rotating plugs with hydraulic locks. The article presents the results of the creation of mathematical models of refueling machine movement and fuel assembly gripping. The time-optimal algorithms for the operation of refueling machines with three rotating plugs are proposed. The use of the new algorithm allows to reduce the time of movement of the grip of the refueling machine by 30–37%.

Conceptual study of subcritical assembly with neutron generator for reactor physics experiments and neutron utilization

In Japan, critical assemblies are regulated under the “Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactions” which provides guidelines for the construction and the operation of nuclear facilities: the core configurations of nuclear experiments are constrained accordingly. In addition, the nuclear facilities are required to comply with strict administrations if the nuclear facilities have more than the certain amount of nuclear fuels. On the other hand, subcritical assemblies are not subjected to the nuclear reactor regulation because the subcritical assemblies are not classified as nuclear reactors. We propose the subcritical assembly with a pulsed accelerator neutron source, which uses nuclear fuels below the regulation amount to ease maintenance and to widen experimental capabilities. Conceptual study of such system was carried out to calculate the deviation between reactivity difference in critical state and that in subcritical state by using the MCNP 6.2 code and the JENDL 4 library. Neutron source multiplication method and pulsed neutron source method were performed. The deviation can be estimated around 10 % by setting detectors in appropriate positions. Additionally, neutron radiography was studied as a simple application. The geometrical shapes of sample rods were seen clearly.

Actual problems of modeling thermal-hydraulic processes in fast neutron reactors

The results of studies on hydrodynamics and heat transfer processes in fast neutron reactors are presented. Data on turbulent momentum transfer in rod bundles are analyzed. It is shown that the intensification of turbulent momentum transfer in the rod bundle channels is due to large-scale turbulent momentum transfer (secondary currents). The intensification of interchannel turbulent exchange in close-packed lattices of rods is explained. A dependence is obtained for the dissimilarity coefficients of interchannel convective exchange forced by wire wrapping in bundles of rods. The methods and results of numerical modeling of thermal hydraulics using the Monte Carlo method, thermomechanical analysis of the temperature field in fuel rod assemblies in the lifetime process are presented. The results of modeling based on a water model of temperature fields and the structure of coolant movement in the primary circuit of the reactor in various regimes are presented. A stable temperature stratification of the coolant was revealed in the peripheral zone of the upper chamber of the reactor above the side screens. It is shown that the process of boiling liquid metals in fuel assemblies has a complex structure, characterized by stable and pulsating regimes and a heat transfer crisis. The agreement between the results of experimental and numerical modeling is shown. A cartogram of the flow regimes of a two-phase flow of liquid metals in fuel rod assemblies has been plotted. The influence of the surface roughness of fuel elements on the boiling process and heat transfer during boiling of liquid metals is analyzed. Long-term cooling of a fuel assembly with a “sodium cavity” above the reactor core in accident regimes with boiling of liquid metals is shown. The objectives of further research are formulated.

Measurement of γ-Rays Generated by Neutron Interaction with 16O at 30 MeV and 250 MeV

Abstract Deep understanding of $\gamma$-ray production from the fast neutron reaction in water is crucial for various physics studies at large-scale water Cherenkov detectors. We performed test experiments using quasi-mono energetic neutron beams ($E_n = 30$ and 250 MeV) at Osaka University’s Research Center for Nuclear Physics to measure $\gamma$-rays originating from the neutron–oxygen reaction with a high-purity germanium detector. Multiple $\gamma$-ray peaks which are expected to be from excited nuclei after the neutron–oxygen reaction were successfully observed. We measured the neutron beam flux using an organic liquid scintillator for the cross section measurement. With a spectral fitting analysis based on the tailored $\gamma$-ray signal and background templates, we measured cross sections for each observed $\gamma$-ray component. The results will be useful to validate neutron models employed in ongoing and future water Cherenkov experiments.

Prompt gamma rays of lanthanum and praseodymium produced by inelastic scattering of fission neutrons

AbstractEmission of prompt gamma rays in lanthanum and praseodymium nuclei triggered by (n,n’γ) inelastic scattering reactions of fission neutrons was investigated with the instrument FaNGaS (Fast Neutron-induced Gamma-ray Spectrometry) at Heinz Maier-Leibnitz Zentrum (MLZ). We identified 125 gamma lines (54 for lanthanum and 71 for praseodymium), for which we give the relative intensities and production cross sections. Presence of oxygen and chlorine in the samples was exploited to verify previous measurements. Our results are consistent with available literature data but also enhance it as we detect new lines and recognize a few false assignments. In addition, for a counting time of 12 h we estimated the detection limits of lanthanum and praseodymium as 0.6 and 0.4 mg, respectively.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 235U

In the epi-cadmium neutron induced fission of 235U, independent isomeric yield ratios (IR) of fission products 130,132Sb, 131,133Te, 134,136I, 135Xe and 138Cs have been measured by using an off-line gamma-ray spectrometric technique. The average neutron energy of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, the root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. The IR and JRMS values of considered fission products in the epi-cadmium neutron induced fission of 235U were compared with the literature data in the thermal neutron induced fission of 235U to examine the influence of excitation energy on nuclear structure effect.

Verification of plastic strain values during ovalization of a ring specimen from a fuel element shell of a fast neutron reactor

Mechanical testing of ring specimens for radial compression between flat dies (ovalization) was carried out, supplemented by analyzing the stress-strain state calculated by a specialized computer program. To verify the values of plastic strain, laser marks were made on the side surface of the ring. It is found that the maximum plastic deformation accumulates at the inner and outer walls of the ring, where maximum tensile and compressive stresses act, respectively. The deviation of the experimental values of plastic strain from the predicted ones does not exceed 10%. It is shown that the analysis of the stress-strain state of the ring specimen during ovalization can be useful for evaluating the critical values of the mechanical characteristics of the shell material irradiated in a fast neutron reactor to a high damaging dose.

Microjet printing of metal salt or oxide targets for nuclear reaction studies on radionuclides

Radioactive targets for direct measurements of neutron-induced reactions are required to improve evaluated cross-section data and, ultimately, the fidelity of neutron reaction network simulations. Electrodeposition and molecular plating are the current state-of-the-art radioactive target production techniques, but not all metals can be electrodeposited or molecular plated with high yields. Alternative techniques can be expensive or may produce targets that are unsatisfactory in terms of thickness, yield, or purity. Microjet printing is a new, rather inexpensive technique that utilizes equipment with a small footprint and has the potential to produce thin, highly radioactive targets with good uniformity and minimal impurities from the target fabrication process. This study involved optimizing the process of producing microjet printed targets to allow for the fabrication of a stable vanadium(V) oxide (V2O5) target that was compared with an analogous electrodeposition V2O5 target manufactured via the application of a vanadium chemical conversion coating on aluminum (Al) foil. The results from these studies suggest microjet printing could be used to produce relatively uniform target layers with adequate film thicknesses (<15μm). V2O5 microjet printed targets, when compared with chemical conversion coating V2O5 targets, appeared to be qualitatively less uniform and quantitatively larger in thickness (11.7(27) μm vs. 5.3(18) μm). However, the conversion coating target contained more impurities and the Al backing had a higher background contribution to measurements of neutron-induced reactions as opposed to targets produced via microjet printing. Overall, the results from this study suggest microjet printing has the capability to be an excellent alternative target production technique to the electrodeposition and molecular plating methods.

A self-organized sandwich structure of chromium nitride for ultra-long lifetime in liquid sodium.

The development of fast neutron reactors with improved efficiency and sustainability, being a tangible solution to the large-scale utilization of nuclear energy, serves as a critical step prior to the commercialization of fusion energy. These reactors use liquid metal coolants, which can weaken the durability of metallic components. Conventional design of protective coatings counts upon thermodynamics, which often overlooks the kinetic factors such as structural evolutions, resulting in deteriorated coating properties. Herein, we present a novel interface-engineering strategy involving the control of the phase transformation direction and interface diffusion reaction. Through iterations of self-organization, desired surfaces and interfaces can be achieved for materials used in harsh environments. Specifically, a CrN-coated steel sample with an interfacial Cr layer was designed and fabricated. After ultra-long (up to 6000 h) immersion in liquid sodium, the CrN/Cr coating structure was converted into a sandwich Cr2N/CrN/Cr2N structure dynamically. As a consequence, the coating system exhibited enhanced properties, namely increased surface hardness (by ∼36%), reduced coefficient of friction (by ∼13%), and enhanced interfacial adhesion (by ∼37%). Thus, the proposed strategy can guide the future design of robust coatings with ultra-long service life in harsh environments.

Simulation Analysis Based on Sodium-Cooled Fast Reactor Fuel Assembly Under Blockage Accident

This paper employs computational fluid dynamics to investigate the fuel assemblies within a sodium-cooled fast neutron reactor. To begin with, we developed computational models for seven wire spacer fuel rods under both normal operating conditions and transient blockage conditions. We conducted separate analyses to assess the impacts of normal operation, blockage thickness, and blockage area. This work allowed us to acquire data on the heat transfer properties and the flow field variations for the coolant, blockages, and fuel rods across different conditions. Subsequently, we leveraged these flow field alterations to examine the resulting temperature distributions. Analyses were conducted to evaluate the effects of normal operation, blockage thickness, and blockage area. The research acquired the heat transfer characteristics and flow field distribution variations of the coolant, blockages, and fuel rods under different operational conditions and utilized these variations to analyze the temperature distribution. Through research analysis, the following conclusions have been reached. Under normal operating conditions, the temperature and flow fields of the fuel components exhibit cyclic variations along the axial length, corresponding to the pitch of the wire spacers. Heat exchange between the internal and external subchannels occurs independently, which further substantiates that the incorporation of wire spacers strengthens lateral flow disturbances. This effect is more pronounced within the internal subchannels, thereby leading to a marked difference in the flow fields on either side of the wire spacers. In the case of blocked conditions, an increase in blockage thickness and area both lead to higher temperatures for the coolant, the blockages themselves, and the fuel rods. The temperature in the recirculation zone behind the blockages also rises with increasing blockage thickness and area, although the magnitude of this increase is not significant. After the onset of blockage conditions, the instantaneous temperature tends to increase. As time progresses, the instantaneous temperature fields following the blockage do not display greater fluctuations in temperature change than those observed after the system has stabilized. For the temperature parameters of the convective field, differences arising from an increase in blockage area are more significant than those caused by an increase in blockage thickness. When blockage area and blockage thickness are increased by the same multiple, an increase in blockage area results in a higher temperature peak. Increasing the area of blockage necessitates a longer duration for both the temperature and the velocity fields to revert to equilibrium.

Activation analysis of RFT-30 cyclotron target assembly

This work evaluated the validity of computer simulation codes by comparing the results of an inventory calculation for a89Zr target assembly with measured data obtained using a high-purity germanium detector. For the inventory calculation, the MCNP6 and FISAPCT-Ⅱ codes were used, and for the measurement, efficiency derived from the efficiency transfer method was used. The MCNP6's ENDF/B-Ⅶ library (TENDL-2017 library for 89Y) and the CEM03.03 physics model were employed to simulate proton reactions, while the ENDF/B-Ⅶ library alone was utilized for neutron reactions. To verify the efficiency obtained by the efficiency transfer method, the calculated activity of the reference sources was compared with the measured activity and confirmed to be within 11%. For the target assembly, the average calculated-to-measured activity (C/M) ratio was determined to be 0.7–2.0 (physics model) and 0.7–1.5 (data library with TENDL), thus confirming the validity of the computer simulation codes. The above results suggest that the methods presented in this study could be used to establish decommissioning plans for the cyclotron facility, including plans for waste disposal of cyclotron components such as target assemblies.

Computational Optimization of 133m Xe Production in the Washington State University TRIGA Reactor

Here we present a new method of irradiating 132Xe capsules with neutrons to produce 133mXe gas standards that are used for radiation detector calibration at radioxenon measurement laboratories in support of the Comprehensive Test Ban Treaty (CTBT). This method is designed to maximize the production of 133mXe compared to 133Xe, both of which are competing products from the 132Xe(n, g) reaction. The 133mXe is produced at a much higher fraction for high-energy neutron absorptions in 132Xe (~50% for fast neutrons versus ~11% for thermal neutrons). We performed “spectral tuning” of the Washington State University (WSU) TRIGA reactor neutron spectrum inside the 132Xe ampules to maximize the number of fast neutrons and minimize the number of thermal neutrons available for 132Xe absorption. Spectral tuning analysis, done with Monte Carlo simulations, provided valuable insights into a future final design for a 132Xe irradiation capsule. With no spectral tuning, the fractional yield of 133mXe in the WSU reactor was ~11.7%. By surrounding the 132Xe capsule with a 0.5-cm-thick layer of tungsten and a 2.83-cm layer of europium (III) oxide and placing it in the reactor’s cadmium rotator tube next to the fuel elements, the fractional yield of 133mXe can be increased to 24.6%, a 111% increase in yield. Thus, by improving the fractional yield of 133mXe through spectral tuning, the CTBT will have better quality gas standards to use for radioxenon detector calibration to assist in the CTBT’s mission.

Evaluating hybrid nuclear-thermal power systems for reduced carbon footprint: A comparative analysis

The article provides an estimated substantiation of the concept of expanding the fuel and energy base and reducing the volume of greenhouse gas emissions with the joint use of organic and nuclear fuel in the structure of hybrid nuclear-thermal power plants. It is shown that the technological scheme (the production of saturated steam in a nuclear steam-producing plant - NPPU, and then overheating to the maximum achieved parameters in the thermal power industry due to organic fuel) leads to a decrease in the specific consumption of gas fuel for the production of electrical energy by 25–30 %, emissions of combustion products by 2.5 times (compared to thermal power plants), consumption of industrial water (compared to nuclear power plants) - by 1.5 times. It is indicated that it is possible to expand the fuel base of modern nuclear energy with thermal neutron reactors through the use of technologies mastered (breeder reactors) and promising (hybrid thermonuclear reactors). The fundamental advantage of the considered complex - JPPP-TPP - its feasibility in a short period of time with a significant reduction in the carbon footprint per unit of energy produced, which is an important condition for the sustainable growth of energy production while maintaining the ecological balance in the geosphere.

Probing earth-bound dark matter with nuclear reactors

Strongly-interacting dark matter can be accumulated in large quantities inside the Earth, and for dark matter particles in a few GeV mass range, it can exist in large quantities near the Earth’s surface. We investigate the constraints imposed on such dark matter properties by its upscattering by fast neutrons in nuclear reactors with subsequent scattering in nearby well-shielded dark matter detectors, schemes which are already used for searches of the coherent reactor neutrino scattering. We find that the existing experiments cover new parameter space on the spin-dependent interaction between dark matter and the nucleon. Similar experiments performed with research reactors, and lesser amount of shielding, may provide additional sensitivity to strongly-interacting dark matter.

Solving a Novel System of Time-Dependent Nuclear Reactor Equations of Fractional Order

Building upon the previous research that solved neutron diffusion equations in simplified slab geometry, this study advances the field by addressing the more complex cylindrical geometry, focusing on neutron diffusion equations that are coupled with delayed neutrons in cylindrical reactors of fractional order. The method of solving used integrates the technique of residual power series (RPS) with the Laplace transform (LT) method. Anomalous neutron behavior is explained by examining the non-Gaussian scenario with various fractional parameters α. The LRPSM Laplace transform and residual power series method employed in this approach eliminates the complex difficulties. This simplicity makes the method particularly coherent with different fractional calculus applications. To validate the proposed method, numerical simulations are conducted with two different initial conditions representing distinct scenarios. The obtained results are presented in suitable tables and figures. It should be emphasized that this system is solved for the first time utilizing fractional calculus techniques. The outcomes are consistent with those achieved using the Adomian decomposition method.

A multi-scale and multi-physical coupling method for the transient characteristics of space nuclear reactor

In the future of deep space exploration, space nuclear reactors (SNRs) are essential because they can provide a robust, sustainable energy supply independent of the environment. SNRs have complex multi-physics coupling effects, and their systems involve many modules. The traditional method uses the lumped parameter model, or single-filed, which cannot accurately describe the transient characteristics and has low spatial resolution and poor data representation. Thus, developing a multi-dimensional, multi-physics coupled method to analyze and design SNRs is necessary. This work proposes a multi-dimensional, multi-physics coupled method for the SNRs. A transient coupling code for the space thermionic reactor TOPAZ-II based on OpenFOAM was developed. The reactor neutron physics model, multi-region heat transfer model, thermoelectric conversion model, electromagnetic pump model, and heat pipe radiator model are included. Different scale models are imported by the code, including a three-dimensional reactor model and a two-dimensional heat pipe. Furthermore, the experimental data from the V-71 test unit were used to verify the developed code. The shutdown and startup of the improved TOPAZ-II system are simulated using the developed code. The results show that the calculated results agree well with the experimental data. During the startup and shutdown processes, the material temperatures remained within acceptable limits, so it can be considered that the developed 2D/3D system analysis for SNRs has successfully achieved a high-dimensional simulation of transient characteristics, accurately capturing the spatial distribution of SNRs at different scales.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 239Pu

Measurement of independent isomeric yield ratios (IR) of 128,130,132Sb, 131,133Te, 132.134,136I, 135Xe and 138Cs have been carried out for the first time in the epi-cadmium neutron induced fission of 239Pu by using an off-line gamma-ray spectrometric technique. The average neutron energy (<En>) of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. Effect of nuclear structure on JRMS values was examined. The present data in the epi-cadmium neutron induced fission of 239Pu were compared with the similar data in the thermal neutron induced fission of 239Pu to examine the role of excitation energy on JRMS values.

Research on a Neutron Detector with a Boron-Lined Multilayer Converter

3He is a splendid neutron detection material due to its high neutron reaction cross section, gaseous state, and nonelectronegative and nonpoisonous nature. With the worldwide problem of the “3He supply crisis” arising, boron-lined gaseous neutron detectors are being widely used in neutron detection to replace 3He neutron detectors. In this work, to reduce the scattering neutron background coming from the substrate of a boron-lined neutron detector in the application of neutron scattering, a new design of the boron-lined gaseous neutron detector composed of a boron-lined multichip converter and a multiwire proportional chamber was proposed. The electron drift efficiency matrix simulated by Garfield++ (Version 2023.4) and the values and positions of electron energy deposition simulated by Geant4 were obtained. The α, 7Li, and total charged particle energy deposition spectra were acquired via coupling calculations of the electron drift efficiency matrix and the values and positions of electron energy deposition, and the width of the slit was selected as 3 mm. The boron-lined multilayer converter neutron detector (BMCND) was tested using a 241Am–239Pu mixture α source, and the total count rate of α charged particles was measured as 599.5 s−1, which is 89% of the theoretical α particle emission rate of 672.9 s−1. The drift voltage experiments showed that 1200 V is enough to acquire a relatively ideal count, and a 2500 V drift voltage was confirmed, considering the higher count and instrument safety. We also performed the neutron detection experiments using a photo-neutron source, and a characteristic spectrum shape of “two stairs” was measured. When borated polyethylene was used to shield the BMCND, the detected total count decreased while keeping the characteristic spectrum shape, demonstrating that the BMCND was equipped with the ability to detect neurons, indicating that BMCNDs have the potential to be an outstanding 3He alternative neutron detector.

Investigation of Cs27LiYCl6:Ce scintillator energy response for D-D fusion neutron spectrometer

During plasma experiments on EAST (Experimental Advanced Superconducting Tokamak), the discharges heated with neutral beam injection (NBI) for plasma results in the production of 2.45 MeV deuterium-deuterium (D-D) fusion neutrons. In order to measure the energy spectrum of D-D fusion neutrons, a new Cs27LiYCl6:Ce (CLYC-7) crystal with 7Li-enrichment has been developed and fabricated into a CLYC-7 scintillation detector. Before utilizing this detector for studying neutron energy spectra, it is crucial to explore the response function of the detector within a relevant neutron energy range. Therefore, the detector has been experimented using the 4.5 MeV electrostatic accelerator at Peking University to investigate its response to fast neutrons, energy resolution, and pulse discrimination capability in the neutron energy range 2.07 MeV–5.51 MeV. The neutron reactions occurring within the CLYC-7 scintillator were simulated using MCNPX, focusing on the reactions 35Cl(n,p) and 35Cl(n,α), as well as the energy response curve and the contributions of different particles (protons, alphas) to the neutron spectrum. The experimental results were successfully validated on EAST, where the 2.45 MeV fusion neutron energy spectrum was measured. That demonstrates the ability of the CLYC-7 scintillation detector to diagnose the fusion neutron energy spectrum for deuterium plasma discharge.

Passive single-sphere multi-detector spectrometry system for profiling of neutron fields in the thermal to 20 MeV energy range

This study explores the development and application of a custom-designed single-sphere multi-detector spectrometer (SSMDSS) for characterizing neutron fields. Stray neutrons in reactor and accelerator facilities present complex, energy-diverse fields. Accurately assessing these fields is crucial for personnel safety. The SSMDSS system is designed to address this challenge, offering a compact and passive dosimetry solution for constrained workspaces. We studied the performance of SSMDSS in two environments: radio-nuclide based neutron sources and stray neutron fields at a research reactor hall. Our investigation demonstrates that SSMDSS provide reliable neutron energy distributions. The performance of the SSMDSS was compared with commercially available ROSPEC system. SSMDSS exhibits superior resolution due to its design, leading to a more accurate measurement of low-energy fluence when compared with ROSPEC. This distinction is reflected in various measured parameters, including average energy and fluence-to-dose conversion coefficients. Inside the reactor hall, SSMDSS measured a fluence-to-dose conversion coefficient of 121 pSv.cm2 compared to ROSPEC's 83 pSv.cm2. Additionally, SSMDSS measured a higher average energy (0.25 MeV) compared to ROSPEC (0.14 MeV). While ROSPEC remains a valuable tool for real-time data collection in dynamic fields with high gamma backgrounds, SSMDSS presents itself as a viable alternative for neutron spectrometry in neutron work fields due to its compact design and passive dosimetry capabilities. This makes it particularly suitable for deployment in constrained workspaces with potential stray neutron exposure.