In this work, we propose and design a GaN-based diode with a p-doped GaN (p-GaN) multi-well structure for high efficiency betavoltaic (BV) cells. The short-circuit current density (JSC) and open-circuit voltage (VOC) of the devices were investigated with variations of parameters such as the doping concentration, height, width of the p-GaN well region, well-to-well gap, and number of well regions. The

Most thickness measurement techniques using X-ray radiation are unsuitable in field processes involving fast-moving organic films. Herein, we propose a Compton scattering X-ray radiation method, which probes the light elements in organic materials, and a new simple, non-destructive, and non-contact calibration-free real-time film thickness measurement technique by setting up a bench-top X-ray thickness

The projected radiation dose to the public due to the release into the environment of the Fukushima treated water was assessed using the methodology of the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR). The assessment was carried out for ten key radionuclides (3H, 60Co, 125Sb, 90Sr, 106Ru, 129I, 134Cs, 137Cs, 14C, and 99Tc) in the treated water, and the mean measured

It is important for operators of poly-pipelines in petroleum industry to continuously monitor characteristics of transferred fluid such as its type and amount. To achieve this aim, in this study a dual energy gamma attenuation technique in combination with artificial neural network (ANN) is proposed to simultaneously determine type and amount of four different petroleum by-products. The detection system

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirably high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal

In this study, an electron-scattering device was fabricated to practically use the ultra-high dose rate electron beams for the FLASH preclinical research in Dongnam Institute of Radiological and Medical Sciences. The Dongnam Institute of Radiological and Medical Sciences has been involved in the investigation of linear accelerators for preclinical research and has recently implemented FLASH electron

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an

The construction of STELLA-2 facility is on-going to demonstrate the safety system of PGSFR and to provide comprehensive understanding of transient behavior under DBEs. Considering that most events are single-phase natural circulation flow with slow transient, STELLA-2 was designed with reduced-height of 1/5 length scale. The ratio of volume to surface area in the vessel can relatively increase resulting

A design parameter study is presented for the closed-loop type passive containment cooling system (PCCS) which is equipped with two heat exchangers: one installed at the inside of the containment and the other submerged in the water pool at the outside of the containment. A GOTHIC code model for PCCS performance analyses was set up and the design parameters such as the heat exchanger sizes, locations

Nowadays, it is necessary to accelerate the construction of new power plant in face of rising energy demand in such a way that the electricity will be generated at the lowest cost while reducing emissions caused by that generation. The expansion planning is one of the most important issues in electricity management. Nuclear energy comes forward with the low-carbon technology and increasing competitiveness

This study is to evaluate the seismic capacity of the fire-damaged cabinet facility in a nuclear power plant (NPP). A prototype of an electrical cabinet is modeled using OpenSees for the numerical simulation. To capture the nonlinear behavior of the cabinet, the constitutive law of the material model under the fire environment is considered. The experimental record from the impact hammer test is extracted

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since

The T-junction exists widely in industrial engineering, especially in nuclear power plants, which plays an important part in nuclear power reactor thermal-hydraulics. However, the existing prediction models of the liquid entrainment are mainly based on the small branches or small breaks while there are a few researches for large branches(d/D>0.2). Referring to the classical models about the onset of

The process of automation and monitoring in industrial control system involves the use of many types of sensors. A programmable logic controller plays an important role in the automation of the different processes in the power plant system. The major control units are boiler for temperature and pressure, turbine for speed of motor, generator for voltage, conveyer belt for fuel. The power plant units

This paper examines sorption of Pd(II) onto illite, MX-80 bentonite, and Queenston shale in Na-Ca-Cl solutions of varying ionic strength (IS) from 0.01 – 6.0 mol/L (M) and pHc ranging from 3 – 9 under atmospheric conditions. A 2-site protolysis non-electrostatic surface complexation and cation exchange model was applied to the Pd sorption onto illite and MX-80 using PHREEQC, and the model results were

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors

The nuclear supply chain attack surface is a large, complex network of interconnected stakeholders and activities. The global economy has widened and deepened the supply chain, resulting in larger numbers of geographically dispersed locations and increased difficulty ensuring the authenticity and security of critical digital assets. Although the nuclear industry has made significant strides in securing

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Analysis Activation) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate

During the postulated severe accident of nuclear reactor, eutectic reaction leads to low-temperature melting of fuel cladding and early failure of core structure. In order to model eutectic melting with the moving particle semi-implicit (MPS) method, the eutectic reaction model is developed to simulate the eutectic reaction phenomenon. The coupling of mass diffusion and phase diagram is applied to

High density polyethylene (HDPE) pipe has many advantages over metallic pipe, and has been used in non-safety related application for years in some nuclear power plants (NPPs). Recently, HDPE pipe was introduced into safety related applications. The main difference between safety-related and non-safety-related pipes in NPPs is the design method of extra loadings such as gravity, temperature, and earthquake

An engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW) is composed of a disposal canister with spent fuel, a buffer material, a gap-filling material, and a backfill material. As the buffer is located in the empty space between the disposal canisters and the surrounding rock mass, it prevents the inflow of groundwater and retards the spill of radionuclides from the

The International Atomic Energy Agency has developed a tomographic imaging system for accomplishing the total fuel rod-by-rod verification time of fuel assemblies within the order of 1-2 hours, however, there are still limitations for some fuel types. The aim of this study is to develop a deep learning-based de-noising process resulting in increasing the tomographic image acquisition speed of fuel

The performance time of human operators has been recognized as a key aspect of human reliability in socio-complex systems, including nuclear industries. Because of the importance of the time factor, most existing human reliability assessment methods provide ways to quantify human error probabilities (HEPs) that are associated with the performance time. To quantify such kinds of HEPs, it is crucial

The mechanical properties of α-Na3(U0.84(2),Na0.16(2))O4 have been researched using the first-principles calculations combined with the quasi-harmonic Debye model. The obtained lattice parameters agree well with the published experimental data. The results of elastic constants indicate that α-Na3(U0.84(2),Na0.16(2))O4 is mechanically stable. The polycrystalline moduli are predicted. The results show

In this paper, from a review of the size distribution of the bubbles during pool scrubbing obtained from experiments by EPRI, we apply the bubble size distributions to analyses on the decontamination factors of pool scrubbing via I-COSTA (In-Containment Source Term Analysis). We perform sensitivity studies of the bubble size on the various mechanisms of deposition of aerosol particles in pool scrubbing

In general, common cause failures (CCFs) have been modeled with the assumption that components within the same group are symmetric. This assumption reduces the number of parameters required for the CCF probability estimation and allows us to use a parametric model, such as the alpha factor model. Although there are various asymmetric conditions in nuclear power plants (NPPs) to be addressed, the traditional

During severe nuclear power plant (NPP) accidents, a H2/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H2/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology

Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake. Elasto-plastic analysis methods are required, and the factors that affect the elasto-plastic behavior of reactor materials should be taken into account. In this study, a material behavior model was developed that considers the irradiation

A simulation-based approach is proposed to study the protective actions taken by residents during radiological emergencies using cognitive findings. Human perception-based behaviors are not heavily incorporated in the evacuation study for radiological emergencies despite their known importance. This study proposes a generic framework of perception-based behavior simulation, in accordance with the ecological

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies

Procedures play a key role in ensuring safe operation at nuclear power plants (NPPs). Development and maintenance of a large number of procedures reflecting the best knowledge available in all relevant areas is a complex job. This paper introduces a newly developed methodology and the implemented software, called iExtractor, for the extraction of syntactic and semantic information from NPP procedures

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition

The solution of the time-dependent neutron diffusion equation can be approximated using quasi-static methods that factorise the neutronic flux as the product of a time dependent function times a shape function that depends both on space and time. A generalization of this technique is the updated modal method. This strategy assumes that the neutron flux can be decomposed into a sum of amplitudes multiplied

The aim of this work considers a second order point reactor kinetics model based on the P1 approximation of transport theory, called in this work as P1 point reactor model. The P1 point reactor model implicitly considers the time derivative of the neutron source which has not been thus considered previously. The inverse method to calculate the reactivity in nuclear reactors -chosen because its high

Conventional integral effect test facilities were constructed to enable the precise observation of thermal–hydraulic phenomena and reactor behaviors under postulated accident conditions to prove reactor safety. Although these facilities improved the understanding of thermal–hydraulic phenomena and reactor safety, applications of new technologies and their performance tests have been limited owing to

Scattering resonance of medium mass nuclides leads complex spectrum in the fast reactor, which requires thousands of energy groups in the spectrum calculation. When the broad-group cross sections are collapsed, reaction rate cannot be completely conserved. To eliminate the error from energy collapsing, the Super-homogenization method in energy collapsing (ESPH) was employed in the fast reactor code

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail

KHNP recently has obtained the approval for the commercialization of the modified 37-element (or 37 M) fuel bundle and now is loading the 37 M fuel bundles in CANDU-6 reactors in KOREA. One of the main issues for approval was the burnup limit. Due to CANDU design and neutronic characteristics, there was no specific burnup restriction of a fuel bundle. The absence of a burnup limit does not mean that

This paper presents ductile fracture simulation of full-scale cracked pipe for nuclear piping materials using the cohesive zone model (CZM). The main objective of this study is to investigate the applicability of CZM to predict ductile fracture of cracked pipes with various crack shapes and under quasi-static/dynamic loadings. The transferability of the traction-separation (T-S) curve from a small-scale

Supercritical CO2 (S-CO2) Brayton cycle has been applied to various heat sources in recent decades, owing to the characteristics of compact structure and high efficiency. Understanding the research development in this emerging research field is crucial for future study. Thus, a bibliometric approach is employed to analyze the scientific publications of S-CO2 cycle field from 2000 to 2019. In Scopus

The Smoothed Particle Hydrodynamics is one of the most widely used mesh-free numerical method for thermo-fluid dynamics. Due to its Lagrangian nature and simplicity, it is recently gaining popularity in simulating complex physics with large deformations. In this study, the 3D single/two-phase numerical simulations are performed on the Liquid Metal Reactor (LMR) centralized sloshing benchmark experiment

The aim of this work is the analysis, design and hardware implementation of the fractional-order point kinetics (FNPK) model along with its closed-loop controller. The stability and closed-loop control of FNPK models are critical issues. The closed-loop stability of the controller-plant structure is established. Further, the designed PI/PD controllers are implemented in real-time on a DSP processor

In order to reduce risks from the Steam Generator Tube Rupture (SGTR) accident and to meet safety targets, various measures have been analyzed to minimize the amount of fission product (FP) release. In this paper, we propose an introduction of a Containment Filtered Venting System (CFVS) connected to the steam generator secondary side, which can reduce the amount of FP release while minimizing adverse

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions

In this paper, the design stress intensity values for the plate-type fuel assembly for research reactor are presented. Through a tensile test, the material properties of the cladding (aluminum alloy 6061) and structural material (aluminum alloy 6061-T6), in this case the yield and ultimate tensile strengths, Young’s modulus and the elongation, are measured with the temperatures. The empirical equations

In this work, an attempt is made to improve the electrochemical corrosion resistance of Zircaloy-4 by helium ions implantation. For this purpose, the Zircaloy-4 was implanted with 300 keV helium ions of fluences 1×1013, 1×1015, and 1×1016 ions-cm-2 by using Pelletron Accelerator. Electrochemical tests of pristine and ion-implanted samples were performed in NaCl solution and their potentiodynamic polarization

Capsules containing NaCl-MgCl2 salt with 316L stainless steel or alloy N samples were irradiated in the Ohio State University Research Reactor for 21 nonconsecutive hours. A custom irradiation vessel was designed for this purpose, and details on its design and construction are given. Stainless steel samples that were irradiated during exposure had less corrosive attack than samples exposed to the same

Timely fault identification is important for the safe and reliable operation of the electric valve system. Many research works have utilized different data-driven approach for fault diagnosis in complex systems. However, they do not consider specific characteristics of critical control components such as electric valves. This work presents an integrated shallow-deep fault diagnostic model, developed

A data-based model, such as an AAKR model is widely used for monitoring the drifts of sensors in nuclear power plants. However, since a training dataset and a test dataset for a data-based model cannot be constructed with the data from all the possible states, the model uncertainty cannot be good enough to represent the uncertainty of estimations. In fact, the errors of estimation grow much bigger

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical–CO2–cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints

The nuclear reactor coolant pump transferring heat energy inherently brings with it the unsteady flow and inevitably threatens to the safe operation of the pump unit, especially with the pressure pulsation induced by the rotor-stator interaction. In this paper, the characteristics of pressure pulsation of the diffuser in a nuclear reactor coolant pump were investigated by the numerical simulation with

Real Time Simulation (RTS) has long been used in the nuclear power industry for operator training and engineering purposes. And, Online Simulation (OLS) is based on RTS and with connection to the plant information system to acquire the measurement data in real time for calibrating the simulation models and following plant operation, for the purposes of analyzing plant events and providing indicative

The investigation of the utilization of enriched 208Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every

A modern approach to nuclear energy involves reprocessing like transmutations of spent nuclear fuel products to reduce their radiotoxicity and time needed for their storage. For this purpose, they are immobilized in inert matrices made of zirconia and can be "burned" in fast neutron reactor or Accelerator Driven System. These matrices in spherical form can be obtained by sol-gel process. The paper

Nickel base Alloy X-750, which is used as fastener parts in light-water reactor (LWR), has experienced many failures by environmentally assisted cracking (EAC). In order to improve the reliability of passive components for nuclear power plants (NPP's), it is necessary to study the failure mechanism and to predict crack growth behavior by developing a probabilistic failure model. In this study, The

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great