Volume 89, Issue 3

In this work, the boiling water reactor (BWR) was adopted as a physical model to study the influence of PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code. The designed reactor pressure can be simulated through the open area of a valve modified by PI controller. The proportional gain ( k p ) is equal to one in all cases. For the integral gain ( k i ) smaller than or equal to 100 (i.e., k i = 0.1, 1, 10 and 100), a scrammed reactor incurred by the high reactor water level (Level-8) that is due to the large reactor pressure drop does not occur in the simulation. Compared to k i = 1, 10 and 100, for k i = 0.1, the reactor pressure modified by the PI controller is more close to the designed reactor pressure; however, the time to meet the designed reactor pressure is longer. The reason is that k i = 1, 10 and 100 incur a larger overshoot in the modified reactor pressure through the feedback system; although the time to meet the designed reactor pressure can be shortened, the amplitude of the reactor pressure varying with time is obvious. To exactly simulate the reactor pressure of a nuclear power plant under the normal operation, our results can offer MAAP code users an important reference to design the PI controller.

The Probabilistic Safety Assessment (PSA) studies can be used to evaluate the time-dependence of the safety systems availability. The purpose of this work is to analyze the unavailability dynamics of the Emergency Core Cooling System (ECCS). Using the PSA methodology, the ECCS was modeled. In order to show how the unavailability of the system varies as function of time (mission time, test interval and component ageing), sensitivity analyses are carried out. Following these analyses, polynomial functions of ECCS unavailability as function of time have been obtained. These functions allow to estimate the unavailability without modelling other ECCS fault trees being a simplified estimation of the system unavailability. In the same time, taking into account that the ECCS unavailability must be less than 10 −2 , the best time variations can be established to have the unavailability complied with this criterion. Also, the best time variations could be similarly established for any safety systems having the unavailability less than 10 −3 .

After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.

High-temperature bent heat pipes are safety-level heat transfer components used in heat pipe cooled reactor, and their applications are becoming increasingly widespread. In this study, a comprehensive three-dimensional numerical model was established to evaluate the effects of different bending angles and radii on the performance of high-temperature heat pipes. The model solved for the temperature, pressure, and velocity fields in the wall, wick, and evaporator, considers the compressibility effect of the vapor. The study investigated parameters such as vapor pressure, vapor velocity, radial secondary flow field, wall temperature, wick temperature, and equivalent thermal resistance of powder sintered core sodium heat pipes with bending angles of 45°, 90°, 135°, 180°, and a bending radius under 90°. The results showed that when sodium vapor enters the bent pipe, the vapor flow deviates towards the outer side, resulting in higher vapor pressure at the outer side than the vapor pressure at the inner side, and the generation of radial secondary vortex. This leads to higher temperature in the evaporator and adiabatic sections of the bent heat pipe, while lower temperature in the condenser. Fractal patterns are observed in the outer and inner wall temperature curve in the adiabatic section, with lower temperatures on the inner side and higher temperatures on the outer side. The bent heat pipe exhibits higher equivalent thermal resistance than that of a straight heat pipe, and the equivalent thermal resistance increases with increasing bending angle, and the equivalent thermal resistance decreases with increasing bending radius.

The hydraulic and thermals performance of flow in a double-tube heat exchanger with an inner twisted triangle tube of length L  = 1 m has been studied numerically. Equations for energy, turbulence, and Navier–Stokes were used to model the fluid flow and heat transfer in a double-tube heat exchanger with an inner twisted triangle tube. The Reynolds number range was 5000–30,000 for nanofluid (water-Al 2 O 3 ) as a working fluid under a turbulent flow regime. The considered configuration was examined by varying the volume concentration of nanofluid while keeping the twist ratio constant (Tr = 5). Different cases of volume concentration of nanofluid are triggered (0.05 %, 1 %, 2.5 %, and 4 %, respectively). The numerical outcomes of a double-twisted tube heat exchanger with an inner triangle-twisted tube are validated with the obtainable numerical data. The governing equations were solved with ANSYS Fluent 18.2. The findings reveal that the larger volume concentration of nanofluid (4 %) gives a greater Nusselt number and vice versa due to the increased thermal conductivity of the nanofluid in the double twist tube heat exchanger compared with a plain tube. However, the nanofluid Nusselt number grows with the rise of both the Reynolds number and volume concentration ( Ø ), and the best value of the Nusselt number is 210 when Reynolds number = 30,000 and Ø  = 4 %. The effectiveness of the heat exchanger grows with an increase in the concentration of the nanofluid and a reduction in the torsion ratio compared to the normal tube. The effectiveness of the double-tube heat exchanger with an inner twist triangle tube increases to the highest value of 0.38 at Reynolds 30,000 and Ø  = 4 %.

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.

The Steam Generator (SG) is a crucial component of a nuclear power plant. Proper water level control in a nuclear steam generator is of great importance to ensure a sufficient cooling source for the nuclear reactor and to prevent damage to turbine blades. The water level control problem of steam generators has been a leading cause of unexpected shutdowns in nuclear power plants, which must be addressed for plant safety and availability. The control problem is challenging, particularly at low power levels due to shrink and swell phenomena and flow measurement errors. Furthermore, the dynamics of the steam generator vary as the power level changes. Therefore, there is a need to enhance the water level control system of the SG. In this paper, a Gain Scheduled Internal Model Control (IMC) based on the Dynamic Sliding Mode (DSMC) method is developed for the level control problem. The proposed method exhibits the desired dynamic properties throughout the entire output tracking process, independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. The simulation results confirm the improvement in transient response achieved by using the proposed controller.

The digitalization enhances nuclear power plant (NPP) instrumentation and control (I&C) system performance, while simultaneously introducing safety and reliability challenges. To enhance the stability of distributed control system (DCS), and minimize the dangerous failure rate and safe failure rate of safety system, signal quality bits (SQBs) are widely used in NPP DCS to identify the validity of important signals. However, in the past decade or so, numerous unscheduled turbine shutdowns and/or reactor trips have resulted from incorrect SQB design or settings. Therefore, it is necessary to conduct optimization research on the design methods and setting principles of SQBs, establish an optimization method for DCS application software SQB verification and validation (V&V), then minimize DCS faults attributed to SQB. With the aim, this work conducts an in-depth investigation and analysis of the definition, design, setting and transmission of SQB in NPP. A specific optimization method for software V&V to conduct SQB on NPP safety DCS application software is proposed. Twelve concerns in the SQB V&V work are detailed. The reactor trip system Δ T protection logic is utilized as a case to illustrate the detail implementation process of this optimization method. This work provides an engineering technical reference and operational guidance for SQB software V&V, conducive to ensuring the safe and economical operation of NPPs.

FLIBE (LiF–BeF 2 ) is both a nuclear reactor coolant and a solvent for fertile or fissile materials. FLIBE can also dissolve a variety of fissile and fertile materials such as uranium, thorium and plutonium and has the ability to serve as liquid fuel for molten salt reactors (MSRs). It’s very high thermal capacity and chemical stability are among its other valuable properties. In addition, the low atomic weights of lithium, beryllium and to a lesser extent fluorine make FLIBE an effective neutron moderator. In this study, cross-section values were determined by using various level density models (constant temperature + Fermi gas, back-shifted Fermi gas, generalized superfluid, microscopic level density models) for reactions of 19 F(n, α) 16 N, 19 F(n, p) 19 O, 35 Cl(n, α) 32 P, 35 Cl(n, p) 35 S and 79 Br(n, α) 76 As, 79 Br(n, p) 79 Se, 127 I(n, α) 124 Sb, 127 I(n, p) 127 Te using TALYS 1.95 code and these datas are compared with the values in the EXFOR database.

Compacting a material means applying pressure so that the material itself loses its internal pores. It improves the homogeneity of the material itself and increases its density. The pore structure of compacted clay is also an important factor that affects its mechanical properties, permeability and water absorption. In this study, MX-80 bentonite clay was first compacted into round cakes of different densities and then subjected to tomography image analysis technique using micro-computed tomography (micro-CT) at millimeter scale of accuracy for 1.0–2.0 g/cm 3 compacted clay (thickness of 0.3 cm bentonite compacted clay). In this paper, we use the latest commercial specifically for scientific visualization, materials science, non-destructive testing, numerical simulation of chromatographic imaging and other functions of a powerful visualization of the image analyzing method.

99 Tc is a crucial high-yield fission product, characterized by its radiotoxicity, extended half-life, high solubility, and significant mobility in the environment. On the other hand, Re is employed as a substitute for Tc due to its analogous properties and non-radioactive nature. Bentonite is commonly used as a backfill material in repositories for high-level radioactive waste storage; however, it exhibits minimal adsorption capacity for Re(Tc). This study aims to enhance the adsorption capability of bentonite for Re(Tc) through modification with Cetyltrimethylammonium Bromide (CTAB). Modified bentonite exhibits excellent adsorption performance under various conditions, primarily driven by the interaction between the ReO 4 − anion and the pyridinium ring of the CTA + . Its primary adsorption mechanism involves a chemical precipitation reaction between ReO 4 − and CTA + . The adsorption capacity is approximately 30 times that of unmodified bentonite, and the adsorption process is characterized by spontaneous low-temperature chemical monolayer adsorption. Experimental results of the modification indicate that high-temperature and alkaline modification conditions contribute to enhanced adsorption capacity, with an optimal modification time of approximately 30 h and a maximum modifier addition of 250 % of the bentonite’s cation exchange capacity (CEC). Additionally, the incorporation of modified bentonite into natural bentonite significantly inhibits the diffusion behavior of ReO 4 − . These findings highlight the substantial improvement in the adsorption capacity of bentonite for Re(Tc) through batch modification, leading to reduced mobility of ReO 4 − within backfill materials.

In order to establish a universal and safety-compliant post-closure safety assessment technique, it is necessary to develop appropriate models to explain the migration behavior of radioactive materials within the rock system. Advection-dispersion experiments (ADE) have proven successful in designing transport models through a calibration/validation process of breakthrough curves (BTCs). In the present investigation, we employed a dynamic column device to examine the transport of Tritium ( 3 H), Technetium-99 ( 99 Tc), and Strontium-90 ( 90 Sr) in crushed sandstone. Non-reactive transport experiments utilizing conservative tracers enabled us to determine the transport parameters, including retardation factors ( R ) and dispersivity ( α ). Our study focused on investigating the surface complexation model (SCM) using an additive approach coupled with the advection-dispersion equation to simulate the reactive transport behavior of 90 Sr. The results affirm the robustness of our chosen thermodynamic database and modeling approach, emphasizing the criticality of accurately modeling and predicting reactive transport behavior in porous media.

Exploring the migration behavior of radioactive nuclides in the surrounding rock media of the disposal site is the basis for the safety evaluation of radioactive waste disposal sites. In this study, the column experiment was employed to evaluate the nuclide migration behavior in the surrounding rock medium of a near-surface disposal site in China and to investigate the advection–dispersion behavior of tritium (HTO) and plutonium-238 ( 238 Pu) in highly weathered argillaceous shale with different degree of weathering. A reasonable numerical model was selected to fit the experimental breakthrough curves (BTCs) and obtain relevant migration parameters. The results showed that: (1) the breakthrough curves of HTO and 238 Pu in fractured media with varying degrees of weathering exhibited clear “peak forward” and “tail dragging” phenomena; (2) the stream tube model can better fit the BTCs of the nuclides in the highly weathered fractures and obtain the average dispersion coefficient < D >, average distribution coefficient < K d >, and other parameters; (3) the two-region non-equilibrium model can better fit the BTCs of the nuclides in the penetrating fractures and weak weathering fractures and obtain the volumetric water content of the mobile liquid region ( θ m ), distribution coefficient ( K d ), and other parameters.

This work examines the opportunities and obstacles related to the novel idea of transforming nuclear waste into biodiesel in the Indian setting. Given India’s increasing energy needs and the necessity for sustainable options, repurposing nuclear waste for biodiesel production presents a unique strategy. The paper examines the possible advantages of this approach, covering economic, environmental, and technological factors, as well as discussing the associated difficulties, such as safety issues, technical obstacles, and public perception intricacies. The goal is to provide valuable information for future research and development endeavors by examining India’s distinct nuclear waste and biodiesel environment, considering socio-economic aspects, legislative structures, and the changing energy industry. The paper provides a detailed analysis that adds to the ongoing discussion on sustainable energy choices, emphasizing the significance of creative strategies in addressing energy demands and handling nuclear waste efficiently.