Reactor Analysis Research Articles

Numerical Study on Subcooled Boiling Flow and Heat Transfer Characteristics in Helical Cruciform Fuel Assembly

The helical cruciform fuel rod is a new fuel design. Its advantages include a large surface area–to-volume ratio, short thermal conductivity distance, and no need for grid spacers. This new fuel rod can effectively improve the hydraulic performance of nuclear reactors. To study the performance of the helical cruciform fuel assembly, the subcooled boiling flow and heat transfer characteristics of this assembly are analyzed in the present work based on computational fluid dynamics. The results indicate that the temperature distribution of the central rod wall surface in the circumferential direction has inhomogeneity and periodicity. The fluid’s temperature and velocity distribution in the cross section are high in the center and low elsewhere, and the fuel rod’s torsional orientation is compatible with the velocity vector’s direction. The vapor volume fraction on the wall of the center rod of the fuel assembly is the highest, and the vapor volume fraction in the mainstream area is relatively low. This work provides a reference for further research on helical cruciform fuel assemblies in the thermal analysis of nuclear reactors.

Validation of McCARD Monte Carlo Calculation Code Against NAA Experiments Using Rare Earth Element Certified Reference Material Activated at JRTR

Rare earth elements (REEs) are widely used in several high-tech modern industries. Quantification of REEs demonstrates a technical challenge that requires the use of special scientific techniques. These techniques commonly introduce uncertainties that may propagate not only during the sample preparation phase but also in subsequent measurement and analysis phases. The Jordan Atomic Energy Commission is starting up a REE investigation program utilizing many of its available physical and chemical analytical capabilities, one of which is the Neutron Activation Analysis Facility (NAAF) at the Jordan Research and Training Reactor (JRTR). The NAAF provides accurate and relatively quick analytical services for better estimations of REEs in the unknown samples of interest without the need for any chemical processing prior to or after sample activation. In this paper, we present analytical results of routinely conducted instrumental neutron activation analysis (NAA) experiments using REE certified samples. To optimize future NAA investigations, first, a time-dependent Monte Carlo code at the JRTR, named Monte Carlo Code for Advanced Reactor Design and Analysis (McCARD), is herein validated via comparing its calculation results of the activation process against the presented results of the NAA experiments. The certified REE samples were activated in a well-thermalized NAA vertical activation hole within the reactor reflector region. Detailed neutronics and burnup calculations as well as transmutations were performed using McCARD. Ultimately, this study aims to assess the accuracy and reliability of the McCARD calculation models as a part of the effort of setting up a dedicated REE analysis laboratory at the JRTR. This study shows very good agreement between both obtained results—NAA experiment and McCARD calculations—with confidence levels noted to be more than 90% for almost all REE elements.

Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor

This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors.

Study and transient neutronics analysis of a 10kWth space heat pipe reactor based on a new dual protection control scheme with control annulus and movable radial reflector

This paper mainly focusses on a new dual protection control scheme based on the control annulus and movable radial reflector for a 10kWth space heat pipe reactor. Through the radial reflector, reactive coarse adjustment can be achieved, while reactive fine adjustment can be achieved through B4C control annulus. Both control schemes allow for independent control. Firstly, the steady-state neutron analysis method is conducted to evaluate the critical safety performance of different core control schemes when the reactor falls into sand or water. The first control scheme uses a movable radial reflector, which can still enable the reactor to maintain a keff below 0.85 when the reactor is soaked into the sand or water. The second control scheme uses an independent B4C control annulus, which enables the reactor to maintain a keff below 0.95 when the reactor is soaked into the sand or water. Through the dual protection control scheme, it can ensure the critical safety for the reactor when the launch or drop accident occurs. Moreover, a transient neutronics analysis is conducted to evaluate the operating characteristics and transient neutronics performance for the reactor. The results showed that compared to the control scheme of Kilopower, when the positive reactivity is introduced, the transient neutron dynamics parameters of this dual protection control scheme are superior to that of Kilopower. And it showed that the dual protection control scheme provides an improved control design for the further design and optimization of control scheme for the space reactors.

Neutron balance approach for inline critical rod position search calculation in Monte Carlo reactor analysis

The critical rod position search has been one of the issues in the Monte Carlo (MC) reactor analysis. This paper proposes a practical and simple approach for the critical rod position search based on the neutron balance equation and the monotonic relation between the control rod absorption rate and the total rod insertion length. The proposed method was implemented in an inline manner within the McCARD, the MC reactor design code, so that the critical rod position is calculated and updated for each fission source iteration. The numerical results in a typical fast reactor problem demonstrate the promising performance of the proposed method in the critical rod position search calculation. Furthermore, the cycle depletion calculation was performed to show the capability of the McCARD to simulate the rodded operation condition. By addressing the critical rod position search problem through the neutron balance approach, the applicability of the MC code in the advanced reactor design will be expanded.

Development of a human reliability analysis framework for nominal human error probability estimate of the TRIGA research reactor in Thailand

Human reliability analysis (HRA) of nuclear research reactors often encounters a lack of human performance data, a challenge that is also faced by the TRIGA reactor of Thailand having no specific full method or human error database for HRA. To overcome this challenge, in 2023, HRA teams from the Korea Atomic Energy Research Institute (KAERI) and the Thailand Institute of Nuclear Technology (TINT) jointly developed an HRA framework for the TRIGA reactor. In the HRA framework development, the HRA practitioners applied KAERI's three main HRA tools, namely 1) the EMpirical data-Based crew Reliability Assessment and Cognitive Error analysis (EMBRACE) method, 2) the Human Reliability data EXtraction (HuREX) database, and 3) the TAsk COMplexity (TACOM) method. The HRA framework covers the overall process including the classification of documents and human error types as well as nominal human error extraction for estimating nominal human error probabilities. According to the results of the first use of the HRA framework on the TRIGA reactor in Thailand, the HRA framework provided an effective way to improve the procedures and systematically identify nominal human errors of actions in the emergency procedures.

Characterizing thermal runaway of reservoir rocks under electromagnetic irradiation towards hydrogen generation from petroleum reservoirs

Electromagnetic(EM)-assisted catalytic heating presents a novel method for in-situ hydrogen production from petroleum reservoirs. This study delves into the interaction between electromagnetic waves and reservoir rocks, characterizing the fundamentals behind thermal runaway (TR) phenomenon in sandstones and shales. Utilizing a custom microwave reactor and advanced analysis techniques, we identify the microwave-induced thermal runaway phenomenon in San Saba sandstone rocks at ∼ 568 °C and Mancos shale rocks at ∼ 253 °C, emphasizing the role of mineral, elemental compositions, and dielectric properties in these differences. We also identified that chlorite, albite, and illite are major contributors to thermal runaway and the significant reduction in power required for reheating rocks, saving 50.0–66.7% for sandstone and 64.0–80.0% for shale. This work contributes new insights into the occurrence and mechanisms of thermal runaway in reservoir rocks, therefore providing an efficient way for enhancing heating efficiency and reducing energy input for in-situ hydrogen production. This research further de-risks the emerging technology for in-situ hydrogen production from petroleum reservoirs via electromagnetic-assisted catalytic heating.

Analysis of AP1000 Small-Break Loss-of-Coolant Accident Using Reactor Transient Simulator

The Westinghouse Electric Company’s Advanced Passive Reactor (AP1000) is characterized by the incorporation of passive safety systems (PSSs) designed to ensure core cooling during transient events. The assessment of PSSs requires evaluation of their performance through a combination of experiments and simulations employing various thermal-hydraulic codes. In addition, detailed evaluation of PSSs for a specific reactor system transient analysis such as loss-of-coolant-accident analysis supports understanding representative integral effects test facility development and the further evolution model development and assessment process. Developing a reactor system code is a complex and time-consuming process that requires significant engineering expertise and effort. It can take several months to even years to complete in the early stages of reactor system design and analysis. However, this process can be expedited through the use of transient simulator models for similar reactor systems, which can be used for lesson learning and training purposes. This study uses the Personal Computer Transient Analyzer (PCTRAN) code. The main advantage of PCTRAN is its ease of use and ability to run faster than real time. This study presents the results obtained for a small-break loss-of-coolant accident (SBLOCA) for two breaks using the full version (licensed) of PCTRAN. The purpose of this investigation is to evaluate the overall system behavior during the postulated SBLOCA event as well as assess the capability of the PCTRAN code to reproduce the system response during transient events. The obtained results were compared with the Westinghouse NOTRUMP system code. The PCTRAN code proved to be reliable in predicting the qualitative behavior of the system in both transient cases. As for the system response, it was found that it is contingent on the activation time of the PSSs. The differences in reactor coolant system pressure between the two codes were attributed to the critical flow model and simplification of mass and energy balance. Despite PCTRAN’s limitations, it can still provide a reasonable prediction of various reactor parameters such as pressure, mass flow rate, and void fraction during a SBLOCA scenario. It is worth noting that PCTRAN currently employs a bulk approach similar to that of the Modular Accident Analysis Program (MAAP) and MELCOR codes. However, the upcoming version of PCTRAN will include an artificial intelligence–based detection and accident prevention system, as well as different models for different reactor components. Consequently, PCTRAN has the potential to be upgraded to match the system thermal-hydraulic codes of the U.S. Nuclear Regulatory Commission and become more widely used in cybersecurity to safeguard nuclear power plants from cyberattacks.

Modeling and numerical analysis for cylindrical flow-through catalytic membrane reactor with power-law reaction kinetics: Revealing dead-core phenomena

Cylindrical flow-through catalytic membrane reactors, employing porous membranes impregnated with catalysts, offer enhanced selectivity and yield in chemical reactions. This work focuses on the mathematical modeling and numerical analysis of a cylindrical flow-through catalytic membrane reactor. The proposed reactor model addresses a critical gap in research by incorporating realistic geometries of porous catalytic membranes. This model simulates a series of irreversible reactions following power-law kinetics under non-isothermal conditions within the reactor, formulated using a system of non-linear differential equations to account for mass and heat balances.The paper investigates the occurrence of dead zones within the membrane reactor as a result of rapid reactant depletion, a phenomenon that has not been extensively studied in prior literature for cylindrical membrane reactors. Problems with fractional reaction exponents require efficient numerical solvers since conventional iterative solvers encounter difficulties due to the fact that the power-law reaction term with fractional reaction exponent is not differentiable at the vanishing concentration. A novel time-marching scheme specifically designed for the cylindrical catalytic flow-through membrane reactor is developed and applied for simulations to get valuable insights into dead-core phenomena.The effects of dimensionless process parameters, including Thiele modulus, Peclet numbers, reaction exponent, Prater number, and a membrane geometrical parameter on the concentration and temperature profiles, as well as dead zone formation, are extensively investigated. The simulation results demonstrate that these parameters affect the occurrence of dead zones and their size. Finally, the effect of convective flow on the reactor performance indicators is presented.

Analysis of gas-cooled micro modular reactor (MMR) fuel

The gas-cooled micro modular reactor (MMR) is a novel concept with a compact fuel form consisting of thousands of tri-isotropic fuel (TRISO) particles dispersed in a silicon carbide matrix. The MMR fuel has a high volume packing fraction and is horizontally placed within graphite channels. This study demonstrates that the horizontal arrangement of MMR fuel within graphite channels resulted in a non-uniform gap between the fuel and graphite brick, leading to an "off-center" temperature distribution within the fuel, and resulting in differences in TRISO performance even within sub-zones positioned symmetrically in the radial direction. Given the complexity of the fuel structure and the challenges involved in simulation, a multi-scale and multi-physics coupled analysis method of MMR fuel was realized in this work by implementing physical models into the open-source finite element platform MOOSE. Thermal-mechanical analysis and failure analysis within sub-zones of the typical design MMR fuel during normal operation were conducted through simulation. The results indicate that higher-temperature sub-zones exhibit increased internal pressure within TRISO and show greater tensile stress in the Chemical Vapor Deposited Silicon Carbide (CVD-SiC) layer. Regarding fuel failure analysis, this work found limited impact from Pd penetration and kernel migration. Pressure-induced particle failure, closely associated with stress in CVD-SiC, emerges as the primary failure criterion. The probability of fuel failure is relatively low, and interfacial cohesive parameters have an impact on the SiC stress and the fuel failure probability at the highest level.

A study on the applicability of simplified few-group GET (Generalized Equivalence Theory) to cylindrical molten salt fast reactor

One of the essential requirements for molten salt reactor (MSR) design is methodology for analyzing multi-physics phenomena, such as the behavior of liquid fuel. In the research of Molten Salt Fast Reactors (MSFRs), the Neutron Diffusion Equation (NDE) is widely employed. This study introduces a method to enhance the accuracy of neutronic analysis of MSFR using the NDE. Using the simple structure of MSFR and the characteristics of liquid nuclear fuel, it is intended to enable the application of the simple equivalence method, which is difficult to perform on the existing fast reactor. A straightforward yet effective approach named the simplified Generalized Equivalence Theory (simplified GET) is proposed for applying flux-volume-weighted homogenized cross-sections and representative Discontinuity Factor (DF) values obtained at material. This approach, while similar to the General Equivalence Theory (GET) method, significantly simplifies the enhancement of accuracy in reactor analysis, minimizing computational efforts. Our investigation spans from the initial core to the burned core, ensuring the applicability of this simple method throughout the reactor's operational life. The proposed method demonstrates promising results, offering a viable solution to improve the accuracy of NDE-based calculations in MSFRs.

Experimental investigation of quench front propagation during reflooding in 5 × 5 rod bundle channel with axial uniform power distribution

This experimental study delves into the impact of inlet injection velocity, inlet subcooling, and linear power density on the quench front propagation during reflooding in 5x5 rod bundle channel with axial uniform power distribution. The investigation utilized a film boiling test facility constructed by NUclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiaotong University. The findings demonstrate a discernible sensitivity of the quench front to variations in these parameters, wherein higher linear power density, reduced flow rate, and decreased coolant subcooling lead to delayed reflooding. Moreover, the study derived the surface parameters of the heater rod use self-developed Reflooding Experimental Data Analysis code (REDA). The comprehensive experimental data and empirical models generated in this study hold significant potential for enhancing the safety analysis of nuclear reactors.

Performance evaluation and validation of OpenMC code for criticality analysis of an MTR-type research reactor

This paper assesses the performance of OpenMC, an open-source Monte Carlo code for particle transport, through criticality analysis of an MTR-type research reactor. The evaluation involves validation against experimental data from the NUR research reactor's first fresh fuel core and a code-to-code comparison with MCNP5, using a consistent methodology ensuring identical input conditions. Key parameters, including excess reactivity, control rod worths, shutdown margin, critical configurations, kinetics parameters, neutron flux, power distributions, and peaking factors, were examined. Results indicate good agreement between OpenMC simulations, experimental data, and MCNP5 results, with marginal differences well within acceptable ranges. Discrepancies with experimental data are below 5 %, considering measurement accuracy, while code-to-code comparisons reveal relative differences below 1 % in most cases. This study supports OpenMC's reliability for MTR reactor criticality analysis, affirming the validity of the developed NUR reactor model for future applications and highlighting the code's credibility and applicability in the nuclear reactor simulations.

From simulation to reality: CFD-ML-driven structural optimization and experimental analysis of thermal plasma reactors

Thermal plasma reactors offer an environment of high temperature, enthalpy, and reactivity, making them highly efficient for solid waste treatment and promising for clean energy production from municipal and industrial waste. Optimal geometrical parameters of the reactor can enhance waste treatment and reactor performance. This study presents a comprehensive analysis of 11 key geometrical parameters of a thermal plasma reactor. Utilizing CFD Fluent software, numerical simulations were conducted to generate a dataset. Subsequently, a predictive model focusing on the average temperature in the core melt zone was trained using six Machine Learning (ML) algorithms. The Particle Swarm Optimisation (PSO) algorithm optimized the hyperparameters of the Gradient Booster Regression (GBR) model, which was combined with a Genetic Algorithm (GA) to identify the reactor's optimal geometrical parameters. A DC arc plasma torch-solid waste thermal plasma reactor treatment system was established on this basis. The study also explored the effects of gasification coefficient, reaction temperature, and thermal plasma jet mode on system performance. Findings indicate that the PSO-GBR model achieved the highest prediction accuracy, with the temperature in the core reaction zone reaching 3621 K. The deviation between numerical simulations and machine learning predictions was a mere 1.3%. Enhancing syngas yield and energy efficiency is achievable by controlling reaction temperature and increasing the gasification coefficient. A laminar plasma jet mode, at equal power, provides a more effective reaction environment. The accuracy and reliability of the machine learning-driven regression model and optimization results are significant in guiding the optimal design of plasma reactors and advancing waste-to-energy conversion processes.

Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method

Abstract An ingenious design, Traveling Wave Reactor (TWR) has been suggested for full exploitation of uranium resources in next generation of nuclear reactors. This design involves the slow propagation of a nuclear fission wave through a long initially subcritical core and the transmutation of nuclear fuel. It has been widely proven that the initiation and propagation of fission waves are feasible prompting the Terra Power Company to conduct preliminary commercialization studies on this project. In equilibrium state, the shapes of the neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial (or radial) directions. In this case the in situ fissile material production and consumption drives the operation. Only natural or depleted uranium is required for fresh fuel region. However, in equilibrium state, the burning region contains a spectrum of fission products as well as higher actinides whose are not easily available for initial TWR core construction. One solution is based on the use of enriched uranium in the first cycle to ignite the fission wave and then employment of the composition in subsequent cycles up to the equilibrium state. The main objective of this work is the feasibility study of forming a fission wave as well as the neutronic analysis of a Gas-Cooled Traveling Wave Reactor in various cycles up to the equilibrium one. The MCNP Monte Carlo code was utilized for the analysis of criticality and burn-up calculation. In each cycle, the axial fresh fuel loading and spent fuel discharge, were subjected to analysis. Results showed that the equilibrium state can be obtained by using special arrays of absorbers, from the third cycle where the shape of neutron flux and the power density distribution in axial direction remain unchanged.

3D numerical simulation on thermal hydraulic characteristics in integrated sodium pool of PHENIX reactor during dissymmetric test

Temperature distribution of internals under different transient events is important for the design and safety analysis of pool-type sodium-cooled fast reactor (SFR). It is influenced by overall flow and thermal stratification phenomena in the multi-scale sodium pool. Due to the difficulty to perform experiment, three-dimensional (3D) numerical method is usually adopted for obtaining the temperature distribution and analyzing complex phenomena in pool-type SFRs. It needs to be further validated using the published benchmark, such as PHENIX Dissymmetric Test. The simplified model in system code and 3D model of local area in current study cannot capture the complex coupling effect of cold and hot pool during dissymmetric test. In this paper, an integrated 3D model of PHENIX reactor primary system both including cold and hot pools is established. The steady and transient simulation with 1800 s of PHENIX Dissymmetric Test process are carried out. The key parameters of simulation are compared with the available experimental and predicted data, which achieves good agreement. Overall flow and asymmetric temperature distribution phenomena in cold pool, hot pool and local area of IHX inlet are simulated. There is a temperature difference between different sides of IHX inlet during a period after scram, with the maximum value of 80 ℃ approximately at 90 s. This validated numerical method can be further used for the analysis of thermal hydraulic characteristics in other pool-type SFR. The obtained temperature distribution of internals can provide the reference for the stress assess of structures in the engineering design.

A review of technoeconomic benefits of torrefaction pretreatment technology and application in torrefying sawdust

Economic analysis of torrefaction process centers on assessment of the economic feasibility of the production and utilization of torrefied biomass using developed models such as costs of biomass, electricity, labour, investment, transportation etc. to evaluate the cost of biomass torrefaction. The increase in energy usage over the past century has raised concern over the energy insecurity and environmental unsustainability of current fossil fuel utilization; therefore, there is need for energy diversification. An attractive alternative is the biomass. However, the poor performance of raw biomass in energy generation further necessitate development of refined technologies to enhance its performance, particularly at low temperatures between 200–300 ℃. This study therefore reviews the technoeconomic benefits of torrefaction technology and reactors and its application in the pretreatment of sawdust. An overview of torrefaction technology, torrefied product characteristics, and economic analysis of torrefier reactor and torrefaction cost/ton were reviewed. From the review, torrefaction significantly improved the physical, combustion and performance characteristics of torrefied products, with comparable durability and storability to raw biomass. Comparing with other thermal pretreatment methods, torrefaction is an economical way of improving biomass properties.

Effect of Cooling Pump Transient Power Supply on ETRR-2 Safety Parameters

The primary goals of the engineering design for nuclear reactors involve safeguarding the integrity of the reactor core. Preserving the integrity of the cladding material is especially crucial, as it serves as the initial defense against the potential dangers posed by radioactive materials. In this work, an accident analysis of core cooling pump power transients of different ratios of the nominal pump power in a typical material test reactor is conducted. Phase failure is a very common electrical fault experienced by three-phase motors. Pump power reduction can be initiated due to several causes, like phase failure, voltage reduction, winding failure, and other causes. The nuclear reactor analysis code PARET/ANL version 7.6 is used to carry out these calculations. The accident scenario began with the reactor operating steadily, then experiencing a transient in the core pump power. This caused the core flow rate to decrease and eventually stabilize at a lower level as the pump power decreased. Core cooling pump power variations ratios of 20%, 33.3%, 50%, and 70% of the nominal pump power are considered in this work. The accident analysis is conducted under the availability and unavailability of reactor safety systems. Reactor safety parameters are reported for all cases of the core pump power variations.

An optimization of efficient combined cycle power generation system for fusion power reactor

Fusion power plants can meet the energy demands of the world. The high thermal performance of a power generation system is still a challenge and one of the developing trends with a fusion reactor due to the high outlet temperature. A combined cycle system has been optimized to work at high outlet temperature and pressure of fusion power reactor with thermal power of 2500 MWth and its input parameter has been optimized along with impact of partial load to achieve high thermal performance very first time. The combined cycle based on the concept of two stages of expansion in the closed-Brayton-cycle and two reheaters in the Rankine cycle named Schematic-IV with the higher thermal performance of 58.19% as compared to Schematic-I, II and III has been proposed for a fusion power reactor. The increase in more than one expansion stage and reheat stage is not economical because it increased the thermal performance by about half of performance as compared to one. The effect of isentropic efficiency of compressor and heat rate have been investigated to validate the thermodynamic model and calculations. The optimized thermal performance of the combined cycle is 58.19% at a feed water mass flow rate of 592 kgs−1 and steam outlet temperature of 277 °C in the combined cycle. The heat rate decreased and thermal performance increased with the mass flow rate verifying that the thermodynamic model is correct. The thermodynamic analysis of the combined cycle-based nuclear reactor provides insights for better system thermal performance and reveals the effect of key parameters on the system performance.

Comprehensive economic analysis of a multi-tubular reactor

ABSTRACT Multi-tubular reactors allow for highly selective and mostly highly exothermic catalytic or non-catalytic reactions to occur making them essential in various industrial processes. Our Present work was to develop a generalised rigorous costing method applicable to industrially designed multi-tubular reactors. In this report, a Fischer–Tropsch (F-T) industrial multi-tubular reactor was rigorously costed. Capital costs totalled to R,52,491,606 with material costs attributing the highest cost contribution (48%). The operating cost amounted to R,18,258,067 with cleaning and maintenance attributing the highest cost contribution(67%). The reactor purchase price was evaluated to R,36,201,107. This differed by 41.5% from the purchase price found using a shell and tube heat exchanger costing method. The total annual cost (TAC) assuming a 3-year payback was found to be R,35,755,269.