Reactor Core Research Articles

A GPU-accelerated linear system solution for the Galerkin finite element method applied to neutron diffusion equation

The In-Core Fuel Management or Loading Pattern Optimization (LPO) is the problem of finding an optimal configuration of nuclear fuel assemblies in a reactor core. Over the years Optimization Metaheuristics based on populations have been used successfully for solving the LPO. However, for such methods, thousands of evaluations of candidate loading patterns with reactor physics codes are required. In addition, more precise calculations are also desirable, which in principle may conflict with faster evaluations. For coping with both criteria simultaneously, a code based on the Galerkin Finite Element Method (GFEM) for solving the neutron diffusion equation designed for Graphics Processing Unit (GPU) is proposed in the present work. The resulting program GFEM_GPU has been implemented in Python language with a library for GPU-accelerated computing. A comparison between GFEM_GPU and the CPU version of GFEM regarding three nuclear reactor benchmarks (IAEA-2D, BIBLIS-2D, and ZION-2D) is provided and the accuracy of the results is compared to the Nodal Expansion Method (NEM). The implementation of the GFEM solve-phase with GFEM_GPU provides iterative solutions up to 50% faster than the CPU version depending on the discretization of the problem. This approach presents several advantages, such as a high-level implementation, faster prototyping, and application to LPO problems, while keeping the high accuracy provided by fine mesh methods.

On the Employment of a Chloride or Floride Salt Fuel System in Advanced Molten Salt Reactors, Part 2; Core Inventory, Fuel Burnup, and Salt Clean-Up System

Breed and Burn (B&B) fuel cycle in molten salt reactors (MSRs) qualifies this reactor type as one of the best candidates to be developed for the Gen-IV R&D program. This feature can be approached by employing a closed fuel cycle and application of a molten salt reactor as a spent nuclear fuel burner; the features promise sustainable and clean energy in the future. In this study, a complete package has been developed to calculate core inventory, fuel burnup, and salt clean-up systems of molten salt reactors during their lifetime. To achieve this, the iMAGINE-3BIC package (“iMAGINE 3D-Reg Burnup & Inventory Calculator package”) has been developed in MATLAB R2023a by employing a CINDER90 module of MCNPX 2.7 for burnup-calculation and multi-linear regression method (MLR). The package can estimate the core inventory (concentration of 25 actinides and 245 non-actinides elements) and the burnup of the reactor core during MSR lifetime (up to 100 years) while optimizing the computational resources (time, CPU and RAM), and it can even be hassle-freely executed on standalone PCs in an appropriate time due to its generous database. In addition, the salt clean-up module of the iMAGINE-3BIC package can be employed to evaluate the effects of the salt clean-up system on the above parameters over the MSRs’ lifetime. Finally, the iMAGINE-3BIC package has been applied to an iMAGINE reactor core design (University of Liverpool, UK—chloride-based salt fuel system) and an EVOL reactor core design (CNRS, Grenoble, France, fluoride-based salt fuel system) to evaluate and compare the performance of chloride/fluoride-based salt fuel MSRs from the point of burnup, core inventory, and salt clean-up systems. The results confirm that while a chloride-based salt fuel system has some advantages in less dependency on the salt clean-up system and fewer poisoning elements inventory, the fluoride-based system can achieve higher burnup during the reactor lifetime. The outcome of this study, along with the first part of this article, provides evidence to support the neutronic decision matrix as well as the pros and cons of employing chloride- or fluoride-based fuel systems in MSR cores.

Direct fusion drive based on centrifugal mirror confinement

A concept for direct fusion drive based on centrifugal mirror (CM) confinement of thermonuclear plasmas (DFD-CM) is described. In centrifugal mirrors, electric and magnetic fields are combined to confine the plasma within a rapidly rotating annulus of burning plasma fixed between two mirror magnets. High energy fusion products leave the reactor core at a rate determined by the velocity of plasma rotation and the strength of the mirrors. Those departing through the jet-side mirror deposit their energy in a “warm” plasma that then expands through a magnetic nozzle to deliver jet power in the 100–1000 kW range. Fusion products departing through the power-side mirror are converted to electricity to power the reactor. Moderate thrusts at attractive specific impulses (10,000+ seconds) are possible. Findings are presented on CM reactor dynamics in propulsion applications, to include new insights into the relationship between mirror and centrifugal components of plasma confinement. Additionally, analysis will be presented on reactor operability limits and characterization of viable configurations based on power density, technology constraints, and the ability to self-power. Ongoing research into the physics of the warm plasma will be discussed, to include description of improved fidelity estimates for fusion energy deposition. Finally, considerations for Alfvén’s frozen-in theorem relative to fusion plasmas and magnetic nozzle performance will be outlined.Viable space commerce and future space exploration will require advanced power and propulsion technologies capable of multi-MW power generation with high specific impulse, moderate thrust levels, and low system specific mass. DFD could make a round trip to Mars in 3 months, and to Saturn, in less than 3 years. A DFD-CM powered mission could deliver an orbiter and 4 atmospheric probes to Neptune in four years. A second-generation Interstellar Probe powered by DFD-CM could reach the edge of the heliosphere (1000 astronomical units) in 25 years.

A differential equation for inverse point kinetics

The reactivity that represents the percentage departure of a nuclear reactor from its criticality condition is perhaps the most important individual property since it is directly related to the nuclear reactor’s safe operation. Control systems are used to limit the rate of increase in reactivity. The development of systems for monitoring the reactivity, from the nuclear power data acquired from the plant's onboard electronics, requires a constant improvement of the nuclear kinetic models.The transient situations that emerge in nuclear reactors can be predicted from the changes in the neutron flux and, therefore, it is possible to make a sufficiently accurate prediction about the consequences of the disturbances relating to the magnitude of the neutron flux, which varies over time, by using the integral parameters that represent the nuclear reactor core. The point kinetic equations relate these parameters and, in this way, allow the transient situations that can occur in a nuclear reactor to be studied.The reactivity is usually determined by solving the inverse point kinetic equation, which is an integral equation that depends on the nuclear power variation history. In this paper, we present a method to calculate the reactivity through the solution of an ordinary differential equation that does not depend on the nuclear power variation history, then allows the prompt recovery of the reactimeter in case of loss of the nuclear power data acquisition or malfunction of such equipment (Zarei, 2022; Ansari, 1991).The presented paper is organized as follows: Section 1 contains the basic concepts of point kinetics equations and the integral formalism for the reactivity calculation. The mathematical formalism that leads to the differential equation for the reactivity is presented in Section 2. The prompt recovery of the reactivity using the proposed method is discussed in Section 3. The results are shown in Section 4 and the conclusion is presented in Section 5.

The European GO-VIKING project on flow-induced vibrations: overview and current status

Abstract The interaction between cooling fluid and solid structures (rods, tubes) in nuclear power plants may lead to flow-induced vibrations (FIV), causing material fatigue, fretting wear, and eventually loss of component integrity. This can cause further safety issues as well as substantial standstill costs due to longer or unplanned outages. With the growing computational power, the application of modern 3D numerical simulation tools for the accurate prediction of FIV phenomena is rapidly increasing. In 2022, the GO-VIKING (Gathering expertise On Vibration ImpaKt In Nuclear power Generation) project received a grant within the Horizon Europe research and innovation funding program. Sixteen European and two US partners started their collaboration in the field of FIV experiments and analysis. Over four years, the GO-VIKING project investigates FIV phenomena occurring in nuclear reactor cores and steam generators under single- and two-phase flow conditions. The project’s main objectives are to expand the expertise in the field of FIV through generation of new experimental and high-resolution numerical data; development, improvement, and validation of fluid-structure interaction (FSI) methods for FIV evaluation; training stakeholders in the application of these methods; and synthesizing guidelines for the prediction and assessment of FIV phenomena in nuclear reactors. This paper provides an overview of the GO-VIKING objectives, scientific program, as well as of the main scientific achievements in the first project year.

Reactor core design with practical gadolinia burnable absorbers for soluble boron-free operation in the innovative SMR

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.

Enhanced reactor core design and safety evaluation for the final cycles of Belgian nuclear power plants

The Belgian nuclear power plants, owned and operated by Electrabel, are approaching the end of their operational lifespan, as per their initial design. Tractebel, as the Owner’s Engineer in charge of fuel management for these units, has maximized the energy use of the existing fuel assemblies, either in the core or in the storage pool, to minimize the need for introducing fresh fuel assemblies in the final cycles. Tractebel’s in-house Loading Pattern Optimizer (LPO) helped in determining the optimum core reloads incorporating sufficient safety margins. The safety evaluation of these cores showed some distinctive features compared to prior reloads containing normal batches of fresh assemblies. Therefore, complementary safety evaluations in addition to the standard cycle specific evaluation were performed and approved by the Belgian Safety Authorities. This paper introduces Tractebel’s improved core design tools and processes, outlining the approach to optimize the reactor core design and providing insights into the safety evaluation methodology employed for the final cycles while avoiding the use of fresh fuel assemblies.

Ligand-gated ion channels as potential biomarkers for ADT-mediated cognitive decline in prostate cancer patients.

Men with prostate cancer are at increased risk of developing cognitive decline by the use of second-generation androgen signaling inhibitors. To date, reliable and sensitive biomarkers that could distinguish men at high risk of cognitive dysfunction under androgen deprivation therapy (ADT) have not been characterized. We used high-throughput transcriptional profiling utilizing human prostate cancer cell culture models mimicking ADT, biomarker selection using minimal common oncology data elements-cytoscape, and bioinformatic analyses employing Advaita® iPathwayGuide and DisGeNET for identification of disease-related gene associations. Validation analysis of genes was performed on brain neuronal and glial cells by quantitative real-time polymerase chain reaction assay. Our systematic analysis of androgen deprivation-associated genes involved multiple biological processes, including neuroactive ligand-receptor interaction, axon guidance, cytokine-cytokine receptor interaction, and metabolic and cancer signaling pathways. Genes associated with neuroreceptor ligand interaction, including gamma-aminobutyric acid (GABA) A and B receptors and nuclear core proteins, were identified as top upstream regulators. Functional enrichment and protein-protein interaction network analysis highlighted the role of ligand-gated ion channels (LGICs) and their receptors in cognitive dysfunction. Gene-disease association assigned forgetfulness, intellectual disability, visuospatial deficit, bipolar disorder, and other neurocognitive impairment with upregulation of type-1 angiotensin II receptor, brain-derived neurotrophic factor, GABA type B receptor subunit 2 (GABBR2), GABRA3, GABRA5, GABRB1, glycine receptor beta, glutamate ionotropic receptor N-methyl-D-aspartate receptor (NMDA) type subunit 1, glutamate ionotropic receptor NMDA type subunit 2D, 5-hydroxytryptamine receptor 1D, interferon beta 1, and nuclear receptor subfamily 3 group C member 1 as top differentially expressed genes. Validation studies of brain glial cells, neurons, and patients on ADT demonstrated the association of these genes with cognitive decline. Our findings highlight LGICs as potential biomarkers for ADT-mediated cognitive decline. Further validation of these biomarkers may lead to future practical clinical use.

Utilizing even Plutonium Isotopes as burnable absorbers for controlling the reactivity and power distribution in Pressurized Water Reactors

In this paper, a new approach is proposed for controlling the reactivity and power distribution in pressurized water reactor cores without the implementation of burnable absorbers. This approach depended on the homogeneous mixing of uranium oxide with even plutonium isotopes; Pu-238, Pu-240, and Pu-242. Three Plutonium fuels; (99% UO2 + 1% Pu-238), (99% UO2 + 1% Pu-240), and (98% UO2 + 0.5% Pu-238 + 1% Pu-240 + 0.5% Pu-242) were simulated using the MCNPX 2.7 code. The three plutonium fuels are compared with the other three erbium fuels; (99.9% UO2 + 0.1% 167Er2O3), (99.74% UO2 + 0.26% 167Er2O3), and (99.61% UO2 + 0.39% 167Er2O3), which are taken as reference fuels. The weight percent of UO2, plutonium isotopes and enriched erbium is adjusted to provide the same initial reactivity. This is important for holding a fair comparison between the two fuel types. The simulation demonstrated that even plutonium isotopes, like enriched erbium, can improve reactor safety parameters. It was found that Pu-240 added fuel can suppress the initial reactivity by 27% and extend the fuel cycle by 6.9% compared to the conventional UO2. The reason for fuel cycle extension is due to the production of Pu-241. Meanwhile, Pu-238 and Pu-even added fuels provide suppressing the initial keff by 13% and 35% and shortening the fuel cycle by 2.7% and 21%, respectively compared to the reference UO2 fuel. Moreover, the fuel temperature coefficient (FTC), moderator temperature coefficient (MTC) and void reactivity coefficient (VRC) of Plutonium fuels are found to more negative than those of erbium fuels at the later stages of burnup because of the spectrum hardening resulting from Pu-239 production.

Interactions Between Molten Sodium and Standard Pipe Insulation

Abstract Safety system design and implementation are critical to the operation of any nuclear plant. For sodium-cooled nuclear reactors, hazards external to the reactor core are present in the form of molten sodium that leaks through degraded piping structures. These structures are often clad in high-temperature insulation to preserve the heat needed to keep the sodium molten in the piping. While large sodium leaks are quite noticeable and often result in hazardous fire situations, small leaks of molten sodium are often masked by the shroud of insulation until a large pool of material has collected outside of the failed pipe. This study concentrated on the physical and chemical interactions between molten sodium and standard fiberglass insulation in temperatures ranging from 100 °C to 500 °C. The degradation of the insulation material begins with the volatilization of the organic binder around 250 °C, thereafter the insulation deteriorates at an advanced rate in areas that are in direct contact with the sodium. Chemical profile data were collected for a variety of sample locations that were in contact with the molten sodium, with only a slight increase in the amount of sodium present that can be attributed to the external sodium source. In this way, the molten sodium chemically reacts with the insulation, which accelerates the degradation of the insulation in locations where the sodium is in direct contact with the insulation. These reactions are enhanced by the temperature of the sodium.

Risk-Informing Nuclear Reactor Safety: The Prediction of the Probability of Core Damage Due to Loss of Power and Cooling

Abstract The modern idea of risk-informed decision-making (RIDM) is here critically examined for all existing, advanced and, generation-IV nuclear reactor systems. Motivated by the predictive difficulties of probabilistic risk assessment (PRA) in regard to occurred accidents, it is evident that the real (not hypothetical) consequences of nuclear core damage accidents that must be considered and quantified are the financial losses, infrastructure damages, societal disruptions, and adverse political policies, and not solely the traditional exceedance of regulated radiation release or public exposure limits. With this perspective, a new dynamic analysis is presented for estimating the probability of core damage due to extended loss of power and cooling in a modern nuclear reactor, giving results different from current standardized PRA/PSA analyses. Using existing data, we approach the multitude of different events in a new way: whatever the initial event in the finite event set {external flood, fire, hurricane, ice storm, typhoon, earthquake, cyber attack…}, the fundamental concerns are the consequent loss and nonrestoration of power, and the inadequate cooling of the core. The present proposed quantitative evaluation uses applicable and fully “exchangeable” severe event data for nuclear and nonnuclear systems, including active and passive emergency back-up systems for a wide range of power losses that lead to widespread damage and societal disruption. While not design-specific, this new independent “order-of-magnitude” estimate for the probability of core damage is some two to two hundred times larger than that shown or reported in recent modern and formal PSA/PRA for licensing submissions using generic failure rate data rather than actual severe event data. This new finding suggests greater uncertainties exist than presently assumed for risk-informed decision-making (RIDM), and points to the need for a major reconsideration and updating of risk assessment and regulatory risk-informed approaches for nuclear plant safety evaluation.

Germination, Physiochemical, and Morphology Changes of TR10 Rice Seeds Irradiated to Filtered and Unfiltered Neutron

In 2020, Sabah’s rice production was 117, 846 metric tonnes with a planted area of 40, 446 ha is relatively low compared to other states such as Selangor, which has 36,004 ha less planted area but can produce up to 159,535 metric tonnes more than Sabah. Low rice production in Sabah has contributed to the shortage of supply. Thus, new varieties could be needed that sufficiently produce yields that comparable to the efforts and cost of rice planting. One of the methods is producing new varieties that probably produce high yields using physical mutagen. From the literature available, gamma and neutrons were found to be the most common agents to increase rice production per unit planting area but used they were used separately. Theoretically, using fast neutron irradiation was found to produce a superior mutant. Thus, this paper aims to study the effect of rice seeds irradiated to both gamma and neutron simultaneous, and compare with neutron alone on the germination rates and physiochemical of rice seeds after the irradiations at different dosages. TR10 rice seeds were irradiated simultaneously to gamma and neutrons in the nuclear reactor core using Pneumatic Transfer System–PTS facility, while irradiation to neutrons alone was performed in a neutron chamber using Beam ports–BP facility at Nuclear Malaysia. The germination study shows a linear decrease as the irradiation doses increase in BP and fail to germinate in PTS. Yet, seeds irradiated in PTS showed no hazardous compounds and a remarkable increase in mineral content in seeds, particularly potassium and magnesium at 14 Gy. Similarly, to the shape of the starch granules in both facilities, the starch structure changes after being exposed to 14 Gy in PTS and 35 Gy in BP. As a result, seeds exposed to neutron alone induced considerably more inclusive effects and may raise growth performance and improve rice quality with optimal neutron dosage range from 7 to 14 Gy.

On the Use of the Jacobian-Free Newton Krylov Method to Generate One-Group Discontinuity and Super Homogenization Factors for Full-Core Neutron Diffusion Simulations

Coupled Monte Carlo (MC) and thermal-hydraulic analysis is valuable as a design or reference tool but can be slow, especially when implemented in a Picard iteration. Previous work has developed a novel prediction block to achieve convergence with fewer MC simulations. The prediction works in two stages: (1) a surrogate-like model predicts macroscopic cross sections on the fly and (2) a reduced-order neutronic model predicts the flux response to the updated cross sections. The main challenge with the prediction block is that the reduced-order neutronic model cannot reproduce the spatial flux distribution with high fidelity. This paper investigates the well-established Jacobian-free Newton Krylov (JFNK) method to preserve equivalency between a homogeneous (nodal diffusion) solution and a high-fidelity transport (MC) solution. Instead of performing multiple computationally consuming MC simulations, the nonlinear iterative approach iterates on correction parameters, e.g., assembly discontinuity factors (ADFs) or super homogenization (SPH) factors, using unexpensive nodal solutions. The JFNK approach does not require additional overhead from the MC solver to generate flux tallies. Further, the approach iterates on diffusion solutions produced directly from a desired code, thus ensuring that the parameters are compatible with that code. The approach is applied to correct a nodal diffusion solution for a realistic three-dimensional pressurized water reactor core. The results obtained in the paper show that the method is very successful in reproducing the heterogeneous solution (up to 2.5% difference in assembly flux for SPH and 0.3% for ADFs) without needing to modify the source code of the nodal diffusion solver. In addition, the results show that ADFs yield the best agreement and are also stable (i.e., weakly varying) when thermal-hydraulic fields are perturbed.

Design study of a 450MW thermal Modified CANDLE fast reactor using helium gas as a coolant

The Modified CANDLE (MCANDLE) burnup scheme divides the reactor core into several regions in an axial or radial direction with equal volume. It works like the CANDLE burnup scheme except that, in the CANDLE burnup scheme, the reactor core is divided into three main regions in the axial direction. Namely, spent fuel, burning region and fresh fuel region. In this study, the design of a modular modified CANDLE fast reactor using helium gas as a coolant has been performed. One of the important roles of a modified CANDLE fast reactor is that it can utilize natural uranium as fuel without the need for enrichment or reprocessing. This type of reactor can be used, including in developing countries, without the nuclear proliferation problem. Using helium gas as a coolant gives hope for a fast reactor due to its properties, such as its less neutron absorption, less radioactive, and as helium is an inert gas, prevents chemical reactions with other materials. The study was performed on a rector with a thermal power of 450MWth. The active core was divided into ten regions with equal volume in radial directions. The refueling scheme has been optimized every ten years of burnup to obtain a good reactor design. In the beginning, the fuel was put in the first region after ten years of burnup was moved to the second region, after another ten years of burnup was moved to the third region and so on until the tenth region where the fuel gets out. The neutronic calculation has been performed using the SRAC (Standard Reactor Analysis Code system). The collision probability method (PIJ) was employed for cell burnup calculation, and CITATION was used for the reactor core design with JENDL 4.0 as the nuclear data library.

Neutronic performance of 208Pb-Bi eutectic-cooled fast reactor with uranium nitride fuel (enriched 15N)

Fast reactors (FRs) require more uranium enrichment than LWRs and fuel reprocessing. The solution proposed to address these issues is to use the modified CANDLE burnup scheme on the reactor core. The use of uranium nitride (UN) as fuel was proposed because the UN fuel has several advantages, such as high melting point, high heavy metal density, and large thermal conductivity. However, using UN as a fuel can produce the radioactive isotope 14C via the reaction 14N (n, p)14C. Therefore, in this study, a neutronic analysis was performed in a modified CANDLE fast reactor employing UN and UN (enriched 15N) as fuel using SRAC. The modified CANDLE fast reactor fuelled with UN (99% 15N) has a higher keff than the UN-fuelled reactor with similar parameters. Similarly, the atomic density of fissile nuclides and power density are higher because the 15N isotope has a smaller neutron capture cross-section than the 14N isotope.

Application of data-driven model reduction techniques in reactor neutron field calculations

High-order harmonic techniques can be used to recreate neutron flux distributions in reactor cores using the neutron diffusion equation. However, traditional source iteration and source correction iteration techniques have sluggish convergence rates and protracted calculation periods. The correctness of the implicitly restarted Arnoldi method (IRAM) in resolving the eigenvalue problems of the one-dimensional and two-dimensional neutron diffusion equations was confirmed by computing the benchmark problems SLAB_1D_1G and two-dimensional steady-state TWIGL using IRAM. By integrating Galerkin projection with Proper Orthogonal Decomposition (POD) techniques, a POD-Galerkin reduced-order model was developed and the IRAM model was used as the full-order model. For 14 macroscopic cross-section values, the TWIGL benchmark problem was perturbed within a 20% range. We extracted 100 sample points using the Latin hypercube sampling method, and 70% of the samples were used as the testing set to assess the performance of the reduced-order model The remaining 30% were utilized as the training set to develop the reduced-order model, which was employed to rebuild the TWIGL benchmark problem. The reduced-order model demonstrates good flexibility and can efficiently and accurately forecast the effective multiplication factor and neutron flux distribution in the core. The reduced-order model predicts keff and neutron flux distribution with a high degree of agreement compared to the full-order model. Additionally, the reduced-order model's computation time is only 10.18% of that required by the full-order model.The neutron flux distribution of the steady-state TWIGL benchmark was recreated using the reduced-order model. The obtained results indicate that the reduced-order model can accurately predict the keff and neutron flux distribution of the steady-state TWIGL benchmark.Overall, the proposed technique not only has the potential to accurately project neutron flux distributions in transient settings, but is also relevant for reconstructing neutron flux distributions in steady-state conditions; thus, its applicability is bound to increase in the future.

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

A critical aspect in the design of a nuclear reactor that needs to be considered is thermal analysis. This parameter relates to molten salt circulation, which serves as a fuel, and coolant in Molten Salt Reactor (MSR) core reactors. This study simulates the temperature distribution performance of the MSR core by considering the effect of the number of molten salt channels with inlet temperature variations. Knowing the temperature distribution will predict the characteristics of heat transfer in the form of natural convection for the survival of the reactor. The reactor used a Molten Salt Breeder Reactor (MSBR) design developed by Oak Ridge National Laboratory (ORNL). The Ansys software was used to simulate a 3D model under steady-state conditions. The geometry was divided into two regions: solid (graphite material) and fluid (molten salt). The thickness of the moderator and radius of the molten salt channel were 0.0508 and 0.0208 m, respectively. The moderator height was 1.98 m. The number of molten salt channels varied up to four. The results showed a temperature difference of 9 to 15 K when the channel increased from one to four at 839 K inlet temperature.

Effect of nuclear reactions rates and core boundary mixing on the seismology of red clump stars

Context. Modeling of the stars in the red clump (RC), that is, core helium-burning stars that have gone through a He flash, is challenging because of the uncertainties associated with the physical processes in their core and during the helium flash. By probing the internal stellar structure, asteroseismology allows us to constrain the core properties of RC stars and eventually, to improve our understanding of this evolutionary phase. Aims. We aim to quantify the impact on the seismic properties of the RC stars of the two main core modeling uncertainties: core boundary mixing, and helium-burning nuclear reaction rates. Methods. Using the MESA stellar evolution code, we computed models with different core boundary mixing as well as different 3α and 12C(α, γ)16O nuclear reaction rates. We investigated the impact of these parameters on the period spacing ΔΠ, which is a probe of the region around the core. Results. We find that different core boundary mixing schemes yield significantly different period spacings, with differences of 30 s between the maximum ΔΠ value computed with semiconvection and maximal overshoot. We show that an increased rate of 12C(α, γ)16O lengthens the core helium-burning phase, which extends the range of ΔΠ covered by the models during their evolution. This results in a difference of 10 s between the models computed with a nominal rate and a rate multiplied by 2, which exceeds the observational uncertainties. The effect of changing the 3α reaction rate is comparatively small. Conclusions. The core boundary mixing is the main source of uncertainty in the seismic modeling of RC stars. Moreover, the effect of the 12C(α, γ)16O is non-negligible, even though it is difficult to distinguish from the effect of the mixing. This degeneracy could be seen more frequently in the future in the new seismic data from the PLATO mission and through theoretical constraints from numerical simulations.

Predicting the core thermal hydraulic parameters with a gated recurrent unit model based on the soft attention mechanism

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.

Cryogenic Liquid Rocket and Nuclear Thermal Rocket: A Survey of the Current Technology and a Comparison for Future Mars Missions

The innovation of reusable rockets holds the potential to revolutionize both Mars scientific exploration and commercial tourism. Thus, this paper aims to indicate the most promising propulsion system for the Mars mission. While SpaceX proposed the cryogenic liquid rocket Starship to achieve Mars colonization by 2050, the viable nuclear thermal propulsion has been outlined by rocketry experts for space travel. This paper surveys the current state-of-the-art of both popular propulsion systems, indicating their advantages and disadvantages. Methods to increase the delivery and cost efficiency are also offered, including on-orbit refueling for cryogenic liquid rockets and various nuclear core variants for nuclear thermal rockets. Subsequently, their comparative analysis is delivered in the discussion, focusing on their transportation efficiency and cost efficiency for future mass missions. With faster travel, greater payload, and reduced propellant consumption, nuclear thermal propulsion has shown to be a brighter prospect and potential than cryogenic liquid rocket. Nevertheless, the author believes a new class of bimodal nuclear thermal electric propulsion system is the most promising option, combining the merits and complementing drawbacks of cryogenic liquid propulsion, nuclear thermal propulsion, and nuclear electric propulsion.