Power Reactor Research Articles

Discussing the possibility of using thorium-based fuels as an alternative fuel to uranium dioxide fuel for APR-1400 reactor

This work investigates the possibility of using thorium-based fuels as an alternative fuel for advanced power reactor APR-1400. MCNPX code version 2.7 with cross-section library ENDF.VII has been used to design an APR-1400 fuel assembly. This code has been used to study the neutronic performance of the proposed thorium-based Fuel types (0.944Th, U)O2, (0.955Th, 233U)O2, (0.934Th, rgPu)O2. The fuel burn-up parameters such as infinity multiplication factor (kinf), initial heavy metals concentrations, Minor actinides concentration and fission products concentration have been analyzed during 1500 effective full power days (FPDSs) for thorium-based Fuel types and compared with the common fuel. The analysis of the horizontal thermal power and neutron flux distribution provides valuable insights into the behavior and performance of the suggested Fuel types in the APR-1400 assembly. The analysis of the neutronic results ensures the viability of using the proposed thorium-based fuel types as an alternative fuel to UO2 because they achieved acceptable safety parameter values and provided a good power distribution through the fuel assembly compared to UO2.

Probing Evolution of the Flux-Pinning Landscape in REBCO Coated Conductors Caused by Gamma Irradiation Using DC and AC Magnetometry: A Novel Approach to Tokamak Magnet Material Development

Optimisation of REBCO coated conductor tapes specifically for use in nuclear fusion will help improve the magnet component lifetimes in future tokamak reactor power plants. The focus of this work was exploration of a novel approach to irradiation studies on REBCO tapes, utilising multiple magnetic measurements to probe evolution of the REBCO flux-pinning landscape more deeply than reported in other studies, for the purpose of identifying primary limiting factors affecting performance. Gamma irradiation experiments were conducted, and pre-/post-irradiation results from DC and AC magnetic measurements using a Physical Property Measurement System (PPMS) are discussed. Magnetisation critical current density (Jc) decreased in all samples with increasing dose, except for the silver overlayer-only samples which did not contain artificial pinning centres (APCs), where Jc increased with dose. Removal of the copper stabiliser coupled with the presence of APCs allowed gamma irradiation to induce pinning force maximum peak shifts, from above 14 T before irradiation to below 9 T afterwards. Flux creep rate varied with the evolving pinning landscape, and the degree of Jc degradation directly correlated with creep rate fluctuations post-irradiation. Changes in critical temperature and diamagnetic saturation also corresponded with changes in Jc and flux creep rate. The major conclusion from this study was that minimisation of flux creep rate is the key to maintenance of performance under fusion-relevant operating conditions. Flux creep manifests as problematic AC losses in all high-temperature superconducting machines; therefore, future work will focus on reduction/prevention of the phenomenon to enhance longevity of performance in any application.

Transient behavior of a molten salt fast reactor under two-phase flow conditions with helium bubbling

In the present study, transient behaviors of a chloride molten salt fast reactor without control rods under two-phase flow conditions are analyzed assuming a bypass of the safety protection system. The two-phase flow in the core is caused by helium bubbles injected to remove the gases of the fission product. The void fraction under the rated condition is assumed to be as low as 0.2 %. However, as the fuel flow rate decreases, the void fraction increases inversely proportional to the flow rate and must be considered in any transient analysis. Therefore, analyses with a point reactor kinetics and thermal-hydraulics coupling model have been performed to account for void flow using the CFD code. Comparison between the transient under two-phase flow and the transient under single-phase flow in which the effect of reactivity due to voids is taken into account with the CFD analysis, shows almost the same results, although there are slight differences due to the difference in fuel volume corresponding to the void fraction. When analyzing transients with a system code that has difficulty handling two-component two-phase flow, the void effect has been considered using the above-mentioned single-phase flow analysis. In this way, the entire plant is analyzed using the RELAP5-3D code and the core inlet conditions are passed to the FLUENT code to analyze the transient behavior. The transients analyzed are UTOP, USBO, ULOFF, and ULOHS under the assumption of bypassing the safety protection system. In addition, a transient assuming the loss of helium bubble supply under the decay heat power conditions is analyzed and discussed. The results of the analysis are compared with the results of the same event without voids analyzed using RELAP5-3D. In the case of single-phase flow, the reactor power changes mainly due to temperature feedback. Although the two types of transients have similar appearance, the breakdown of reactivity is quite different. Relating the comparison between the calculated result using the system code and CFD code shows that for the short time behavior until the fuel flow rate shifts to low flow rate, the results of the CFD code, which can analyze the behavior of the helium voids, are more appropriate. However, there are no major problems in the same transient calculation using RELAP5-3D, which calculates the entire plant system.

Secondary system modeling and primary system coupled simulation using MARS-KS

The flexibility of nuclear energy can be achieved by the load following operation and the thermal energy utilization for hydrogen production, desalination, district heating, and Thermal Energy Storage (TES). Thermal energy utilization is performed by transferring the remaining thermal energy over the energy required for electric power generation without changing the reactor power level significantly. For the Pressurized Water Reactor (PWR), it becomes possible by extracting steam from the secondary system and circulating it to a thermal energy application such as the TES system. However, the steam extraction operation affects the behavior of all thermal–hydraulic (TH) variables in the secondary system and further influences the primary system in terms of coolant temperature, reactivity, and reactor power. For thermal energy utilization, it is necessary to secure a technology that can examine the overall stability of the primary and secondary systems in a Nuclear Power Plant (NPP). In this study, based on the system analysis code MARS-KS, an NPP analysis model for SMART, a small-sized integral type PWR, was developed that can simulate major TH variables behavior in the primary and secondary systems at once. Its strength is that major components in the secondary system are modeled without any boundary conditions under matching the heat and mass balance. From the verification calculation of the steady state, it was confirmed that this SMART analysis model can predict the main TH variables of the primary and secondary systems similarly to the plant data. Furthermore, an example calculation was performed for the TES application. The analysis results show that the analysis model is sufficiently applicable to the transient calculation of the primary and secondary systems during the steam extraction operation. This analysis model is expected to be used for the development of technologies such as load following operation and thermal energy utilization in the future.

The role of risk-informed approaches for advanced reactors in Korea

Currently, various types of advanced reactors are being developed in the Republic of Korea, including (1) advanced large power reactors, (2) light water–based small and modular reactors, and (3) Generation IV reactors. While Probabilistic Safety Assessments (PSAs) have been performed for these advanced reactors in Korea, the PSAs were not conducted in detail since they were not regulatory requirements in many cases. However, the role of PSA is changing in Korea nowadays, and qualified PSA models have recently become an essential element in the licensing of operating and new reactors. In addition, PSAs for advanced reactors are essential not only to meet regulatory requirements but also to improve the safety and performance of advanced reactors. It can be said that the success of advanced reactors depends on how their safety and economics can be secured effectively and efficiently. The PSA will play an essential role in this aspect. Various risk-informed approaches (RIAs) based on PSAs are used in many countries worldwide for this purpose. A few issues remain to be resolved for the success of PSAs and RIAs for advanced reactors, such as the development of licensing frameworks for advanced reactors, the reliability assessments of passive safety systems, and so on. In this paper, we introduce the current status of PSAs and RIAs for advanced reactors, including descriptions regarding the characteristics of advanced reactors currently being developed in Korea. We also discuss some important issues to be resolved for the success of RIAs for advanced reactors. We also briefly introduce some research performed to resolve those issues.

Preliminary analysis of startup transient for an ultra-small lithium-cooled space reactor power system

A multi-phases startup scheme based on the reactor startup first, followed by the closed Brayton cycle (CBC) machinery was proposed for the startup condition of an ultra-small lithium-cooled space reactor power system (ULCR) with a CBC conversion system concept scheme. The scheme minimizes the auxiliary motor power required for the turbomachinery startup. The overall startup process includes 7 phases. In the low-power operation phase, by controlling the turbomachinery shaft speed and helium-xenon (He–Xe) working fluid mass flow, the turbomachinery output power was balanced with the needed motor power. Therefore, the auxiliary equipment of the power conversion system was minimized. As the turbomachinery speed and He–Xe mass flow increased to the rated conditions, the core was in low temperature and low power steady-state operation, which was conducive to keep system in safety conditions for the automatic control system. With the reactivity insertion, the core temperature was finally increased to the rated operating condition, and the CBC conversion system's output power increased to the rated output power. The transient process of the designed startup scheme was calculated and analyzed by using the developed transient analysis program TACSNR. The results show that the change rate of the core average temperature at the rising phases was less than 1.1 K s−1. During the whole startup process, the parameters of the ULCR system are all in the safe range.

Study on load tracking characteristics of closed Brayton conversion liquid metal cooled space nuclear power system

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

Research on thermal stratification of surge line in the marine reactor based on fluid-solid thermal coupling model

Due to the requirements of ship movement, the power of the marine reactor is constantly changing, and the expansion and contraction of the primary coolant make the fluid flow in and out through the surge line. This paper establishes a fluid-solid thermal coupling model for the thermal stratification of a Marine reactor surge line. The low Reynolds number eddy viscosity model and the curvature modification of the bending pipe are used in this model. Qiao experiment is provided to verify the model's applicability to the thermal stratification analysis. The typical working conditions of the Marine reactor are analyzed, and the results show that: the temperature difference between the top and bottom of the pipe changes constantly with the fluctuation of the fluid, the maximum temperature difference in the horizontal section is 59.6 K, and the maximum temperature difference in the vertical area is 23.9 K. The maximum stress fluctuation of the horizontal section is 36 MPa, while that of the vertical section is 5 MPa. In the non-destructive testing, many small pits were found in the horizontal section of the pipeline, while no obvious damage was found on the surface of the vertical section, which is consistent with the results of our numerical analysis.

ВВЭР-1000 ЯДРОЛЫҚ РЕАКТОРЫНДАҒЫ ЖЫЛУФИЗИКАЛЫҚ ПРОЦЕСТЕРДІ КОМПЬЮТЕРЛІК МОДЕЛЬДЕУ

The priority use of water-water reactors in the nuclear power industry is due to their advantages. Water is ideally suited as a moderator and refrigerant, providing compactness, high power and simple reactor design. Water-water reactors have high resistance and self-regulation due to the negative temperature coefficient of reactivity. Their activity is caused by short-lived nuclides, which facilitates biological protection and access to equipment. Water effectively removes heat. However, there are also disadvantages: high neutron absorption by water, the requirement of uniform distribution of water, corrosion activity, expensive cleaning system and temperature limitation. An increase in unit power is possible due to the equalization of heat generation in the core. This requires an accurate thermohydraulic calculation. As a result of the study, a theoretical thermohydraulic calculation of heat transfer in the heat - generating set of the VVER-1000 reactor was obtained for the weighted average and maximum values of heat release in the heat-generating set. Computer simulation of heat transfer was also carried out for the distribution of moderately loaded energy in the heat-generating set. The results obtained were compared and analyzed. As a result of the calculation, the distribution graph and temperature field of the heat-releasing element of the coolant, fuel and reactor in the shell of the VVER-1000 heat-releasing kit were obtained.

Conceptual design of a novel megawatt portable nuclear power system by supercritical carbon dioxide brayton cycle coupled with heat pipe cooled reactor

In order to meet realistic requirements for flexible and reliable power supply in special scenarios such as remote areas and deep ocean exploration, a novel megawatt portable nuclear power system called SUPERHERO (SUPERcritical carbon dioxide cycle-HEat pipe ReactOr power system) is proposed in this paper. The system consists of a heat pipe-cooled reactor and a closed supercritical carbon dioxide (S–CO2) Brayton power cycle. Different cycle concepts were compared, and the simple recuperated cycle was adopted for compactness and simplicity of the system. The multi-objective optimization is performed to maximize cycle efficiency and minimize system volume, resulting in optimized parameters for a 1MWe system. By utilizing uranium nitride (UN) modular fuel elements and high heat transfer capacity potassium heat pipes, a compact reactor scheme with 3.13MWt is determined that can operate for 15 years without refueling. The neutronics and thermal analysis of the reactor are carried out, and the results show that the reactor has safety features such as a low peak factor, negative temperature feedback coefficient, sufficient reactivity compensation and control, and anti-failure of single heat pipe. The primary heat exchanger is also designed using the CFD method, and the heat exchange tubes combined with ribbed plates and self-supporting structures are used to enhance heat transfer and ensure structural stability. All these studies outline the general features of SUPERHERO, demonstrating that SUPERHERO has great development potential.

Optimal Stabilization of Reactor Power with Preliminary Identification of Its Characteristics under Conditions of Uncontrolled External Influences

Optimal Stabilization of Reactor Power with Preliminary Identification of Its Characteristics under Conditions of Uncontrolled External Influences

Adaptive neural network sliding mode controller design for load following of nuclear power plant

The power control and load following of nuclear reactors have always been a research topic of concern in the field of nuclear power. In order to improve the control system performance and load following performance, this paper proposes a sliding mode control method based on adaptive RBF neural network (RBFNN) for pressurized water reactor (PWR) power control and load following. Firstly, based on the operation mechanism and energy transport process of PWR, the PWR power model is established based on the point-reactor equation. Then, based on Lyapunov stability theory, an ideal sliding mode controller is designed for a general nonlinear second-order model. Secondly, considering that some state variables in the actual PWR model cannot be directly measured, which leads to the problem that the actual control law cannot be directly applied, RBFNN is used as the controller, and its output is used to replace the actual control law and approximate the ideal control law. Adaptive algorithm is adopted to improve the approximation effect of RBFNN. Finally, the simulation results show that the proposed control method can achieve the PWR load following task well under the set working conditions, and has good robustness to disturbances. Compared with the traditional sliding mode control (TSMC) method, this method can effectively eliminate chattering phenomenon and greatly improve the control performance of the sliding mode controller.

Modeling of power reactors for VFTO simulations: Full-scale design in a real application

The shunt reactors operate under a very specific conditions in Gas Insulated Substations - GIS, especially during very fast electromagnetic transients due to switching and maneuver operations. A meticulous electromagnetic analysis for correct design of such power devices, especially on the insulation coordination between turns, discs, and from winding to ground. This article presents a step-by-step modeling methodology based on the Finite Element Method, which is applied during the conception and manufacturing of a real shunt reactor for GISs. Finally, the time- and frequency-domain simulations are compared with experimental data obtained from laboratory tests.

Preliminary analysis of maximum hypothetical accident for a solid core space nuclear reactor power system

In order to study the safety characteristics of the solid core space nuclear reactor power (SNRP) system under the maximum hypothetical accident, a two-dimensional entire core transient heat transfer analysis model was established, and the key parameters response characteristics of the ultra-small lithium-cooled SNRP under two maximum hypothetical accidents, namely, loss of heat sink (LOHS) and loss of coolant accident (LOCA), were calculated and analyzed. In the maximum hypothetical accidents, the coolant cooling capacity is lost, thus the core decay power is discharged into space only through the radiation of the residual heat removal system and the side surface of the reactor vessel. During the accidents, the heat transfer of the core deteriorated, and the fuel temperature may rise to the melting point, resulting in radioactive leakage. The results show that: (1) In the LOHS accident, the maximum fuel temperature reaches 2150 K at 550 s, and the pressure of the primary system volume accumulator continues to increase to the set system pressure safety limit of 2 MPa, resulting the primary loop overpressure failure. And the fuel matrix temperature is close to the set cladding limit temperature of 2200 K; (2) In the LOCA, the deterioration of heat transfer in the core makes the temperature increase rapidly, reaching a maximum of 3016 K at about 630 s, which is very close to the melting temperature of UN fuel 3123 K. As the decay power decreases, the maximum core temperature decreases to less than 1600 K after 24 h of the accident. The auxiliary cooling system of the solid core SNRP system under the maximum hypothetical accident is optimized, and the design parameters of the auxiliary cooling system meeting the safety requirements are obtained.

A machine learning approach for predicting the reactivity power of hypervalent iodine compounds

The knowledge of chemical reactivity of substrates is a prerequisite to accurately design a chemical reaction; however, it has been a challenging task due to the slow trial-and-error experimental approaches and the high computational cost associated with in silico investigations. Artificial intelligence techniques could serve as an alternative to efficiently determine the relative reactivity of chemical entities. In the context of this research, we propose an artificial neural network model to predict the bond dissociation energies of hypervalent iodine reagents. An open-source cheminformatics package, namely, Mordred, was employed for calculating various 1D, 2D and topological descriptors. The approach utilizes a dataset of more than 1000 hypervalent iodine reagents, and the bond dissociation energies can be predicted with a remarkable accuracy, as suggested by an R2 score of 0.97 and a mean absolute error of 1.96 kcal/mol. Owing to the low cost and high efficiency, this machine learning approach can provide an alternative to the theoretical/experimental approaches to rationally design a chemical reaction and without having to go through the hassle of high-throughput experimentation to reach the desired reaction outcome. In an effort to make the model interpretable, a feature importance algorithm was applied, which identified descriptors contributing most to the development of the model. Features describing electronegativity and polarizability are some of the important contributors to the model’s training.

Research and analysis of the thermal and control characteristics of the plate-type fuel assembly for the supercritical CO2 reactor

Small nuclear reactors can be applied to micro grid power generation, island and space nuclear power systems. The plate-type fuel assembly has a compact design structure, low fuel core temperature, large heat exchange area, and high heat exchange efficiency, which is widely used in small nuclear reactors. To supplement the lack of research in S-CO2 small nuclear reactor, The Brayton cycle reactor system analysis program of S-CO2 plate-type fuel assemblies (BRESA-PFA) is developed, and the reactor control system program is developed to simulate the transient condition in this paper. This article establishes a fuel assembly flow and heat transfer model and a control rod control model using Modelica language, achieving physical and thermal coupling and power control of the reactor. The steady-state operating parameters of the developed program BRESA-PFA were studied, and the flow distribution, coolant and fuel temperature distribution of BRESA-PFA were obtained. The flow distribution conforms to the experimental parameters of CARR, and the fuel assembly radial temperature distribution is high at the center and low at the edge. After verification, the maximum fuel temperature did not exceed the limit value and met the design requirements. Transient characteristic analysis was conducted to study the rod lifting strategy during the startup process, the start-up and rod lifting process of BRESA-PFA is N2-N1-G2-G1, using the logic of step control. Research on reactor control system response under loss of coolant accident, after the accident, the coolant flow suddenly decreased to 65 % of the rated value, and the reactor power also decreased to 65 % FP, G2 and G1 control rods were inserted downwards, the coolant temperature fluctuates by 1 %∼2%. The reactor control system was able to respond quickly, ensuring the stability of coolant temperature, proving the safety and reliability of BRESA-PFA, and providing theoretical reference and scheme support for the research of S-CO2 small nuclear reactors.

Evaluation of uncertainty in antineutrino spectra normalization calculations for advanced nuclear reactor monitoring

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation.

Measurement of Isfahan heavy water zero-power reactor kinetic parameters using advanced pulsed neutron source method in a near critical state

The pulsed neutron source (PNS) technique was used to determine the prompt neutron decay constant for two different lattice pitches in the HWZPR heavy water zero power reactor. The results were compared to the variance-to-mean ratio (VTM) method. The neutron mean generation time was also calculated for both pitches, and the results were compared to previous Monte Carlo calculations. The findings of this research can be used as a benchmark nuclear codes to validate kinetic parameters.

Investigation of reactivity control strategies for a micro-transportable gas-cooled thermal reactor

Micro nuclear reactors have gained increasing attention in recent years due to their high efficiency and long lifetime, making them an attractive solution for meeting the energy needs of remote and off-grid locations. Previous work has proposed a conceptual design for a micro solid nuclear reactor cooled by a helium-xenon mixture. The main aim of this study is to examine the effects of introducing burnable poisons on the neutronic performance of the designed micro nuclear reactor and to improve the design of the sliding reflector segment for reactivity control. The computational results indicate that the inclusion of burnable poisons can decrease the initial excess reactivity and radial power peaking factor, while not having a significant impact on the lifetime of the core. The use of 0.4%wt Gd2O3 and 0.75%wt ZrB2 as burnable poisons results in a reduction of 6.0 $ and 7.4 $ in the maximum reactivity of the core, respectively, while the core lifetime is reduced by 169 EFPD and 545 EFPD, respectively. By partitioning the sliding reflector, a linear introduction of reactivity can be achieved, and the influence of introducing burnable poisons on the reactivity control ability of the sliding reflector is relatively small.

An attempt to find the optimal loading pattern of minor actinides in the PWR fuel rods

Three approaches of minor actinide (MA) loading in the fuel rods of a pressurized water reactor were examined. They comprised inserting MAs and fuel mixture in the center of duplex fuel rods, coating of MAs and fuel mixture in the outer layer of fuel rods and lastly, mixing MAs homogeneously with the entire fuel rod elements. AP1000 fuel assembly was chosen as the reference design. Two fuel types namely conventional UO2 and UTh MOX fuel were selected as the base fuel models for the study. Although shorter cycle length than the UO2 fuel design was achieved by the UTh fuel design, UTh fuel exhibited more favorable FTC and yielded a lower plutonium stockpile at the end of the fuel cycle. When MAs were added in the outer layer of fuel rods, it resulted in higher rate of transmutation, however it had an adverse impact on the fuel reactivity, discharge burnup and pin power distribution. On the contrary, when MAs were loaded in the central region, it led to incomplete incineration of MAs. Therefore, based on the satisfactory transmutation rate, MTC, most negative FTC and uniform pin power distribution, the homogenous MA loading strategy can be concluded as the optimal loading pattern for the AP1000 nuclear reactor.