Due to the demands for an accurate and realistic severe accident code in the PHWR safety analysis, a new severe accident code called CAISER is being developed at Korea Atomic Energy Research Institute (KAERI). The CAISER code simulates core degradation phenomena occurring in a calandria tank, and it consists of two main modules: A fuel rod degradation module and a fuel channel degradation module. The

As one of the promising accident tolerant fuel (ATF) cladding materials, FeCrAl can efficiently reduce the hydrogen generation and heat release under accident conditions due to its excellent oxidation resistance. In the case of postulated accidents, such as the loss of coolant accident (LOCA), coolant is immediately ejected to achieve rapid cooling of reactor core. In order to investigate the liquid-vapour

An issue of estimating the impurity gas amount in fresh ceramic nuclear fuel is discussed from the point of view of their possible influence on parameters of in-pile experiments simulating a severe accident with core melting. A method based on measuring the pressure and temperature of gas in a closed cavity where fuel is fast melted is proposed and experimentally tested under in-pile experiment conditions

Light gas stratification erosion by steam turbulent buoyant jet is the typical phenomenon about containment thermal hydraulics and hydrogen safety. Overprediction of stratification erosion rate has been observed in computational fluid dynamics (CFD) analysis by using current turbulence model based on simple gradient diffusion hypothesis (SGDH). To solve this problem, the effects of buoyancy term, coefficient

The Chebyshev rational approximation method (CRAM) has become a widely adopted method for solving nuclear depletion problems. Therefore, understanding CRAM’s accuracy is important for the safe operation of nuclear power plants. This article performs a global error analysis of CRAM and finds that, as the length of the time step approaches zero, the relative error measured between the exact and CRAM

Global sensitivity analysis (GSA) aims at quantifying the individual effects of input variables on the model outputs. The global sensitivity analysis is expressed by the Sobol’ indices. The Sobol’ indices can be obtained by different algorithms. These algorithms can be divided into two main categories: the sampling methods and the surrogate models. This work employs these methods to perform the global

A performance analysis of a multi-stage safety injection tank (MSIT) was carried out using a numerical method. A MSIT is capable of injecting coolant from a single tank in multiple stages in the event of a loss of coolant accident (LOCA) at a nuclear power plant. The performance and consequences of two- and three-stage MSITs were compared. The effects of break size, installation heights of the injection

The rod ejection accident (REA) is the most severe reactivity-initiated accident of the pressurized water reactors (PWRs). The three-dimensional (3-D) neutron kinetics method is widely applied to the REA analysis, where the fixed neutron source is usually ignored at the hot zero power (HZP) condition. This paper describes a new capability of the 3-D nodal diffusion code, RAST-K to consider the fixed

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th, 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions and core meltdown due to station blackout. Although research efforts have been made by various parties to study the accident scenarios since the Fukushima accident, there remain unresolved issues regarding

In this paper, a benchmark experiment of tungsten granular target worth has been performed on the VENUS-II light water zero-power facility of CIAE in Beijing. The reactivity worth of the tungsten granular target was measured and processed as (-1.007±0.096) mk by a period method. In MCNP simulations, three modeling methods of tungsten grains, i.e., the homogeneous, body center and face center models

The seismic probabilistic safety assessments of nuclear power plants (NPPs) conservatively assume that the correlation between the seismic failure to structures, systems, and components (SSCs) is independent or fully dependent. This, however, is an extreme assumption, and so a more appropriate seismic failure correlation should be considered for accurate safety assessments. Quantification of the correlation

The stellarator plasmas devices have some advantages in steady state operation with less magneto hydrodynamic activities, and the development of stellarator-type FFHR (Fusion-Fission Hybird Reactor) can provide viable energy in the foreseeable future with less requirements to the plasmas technology compared with the pure fusion system. However, the neutronics design of the stellarator-type FFHR has

This paper presents GOTHIC analyses of hydrogen deflagration tests performed in the framework of the OECD/NEA THAI program. The experiments were performed in one or two interconnected vessels under similar conditions, including cases with forced convection loops and both upward- and downward-propagating burns. To capture the influence of initial turbulence and flame wrinkling, a global burn enhancement

Nuclear power plant is a complex engineering system with strong coupling, nonlinearity and potential radioactive release risk. Different researchers have proposed numerous fault diagnosis techniques but there is a gap between these techniques and their practical manifestation. In this paper, kernel principal component analysis (KPCA) and similarity clustering are primarily presented for fault diagnosis

Cold spray technology is a promising solid-state, powder-based deposition methods for fabrication of coatings, near-net-shape manufacturing, and component repair. In recent years, this technology has been investigated for many applications in the nuclear energy sector. For example, it has been explored for the deposition of coatings of corrosion and oxidation materials on zirconium-alloys fuel claddings

In this paper, dryout phenomenon of steam-liquid two-phase flow in vertical tube is studied. Based on Becker’s experimental data, 393 points of dryout data with pressure ranging from 2.9 to 20.1 MPa, mass flow rate ranging from 496 to 3120 kg/(m2 s), heat flux less than 1268 kW/m2 and pipe diameter ranging from 10 to 24.7 mm are re-fitted, and a new correlation for predicting the critical quality is

This study mainly investigated the Bi2O3-B2O3-ZnO ternary glass to immobilize the iodine waste forms under low sintering temperatures (≤600 °C). Based on the results of Molecular Dynamic (MD) code GULP and amorphous fraction, the Bi2O3-B2O3-ZnO ternary glass with 20 mol% Bi2O3 to be chosen to immobilize the sliver-coated silica gel with 10 wt%, 20 wt% and 30 wt% iodine loadings (total content of iodine

Gamma-ray emission tomography (GET) was developed as a potential verification tool to visualize the passive gamma-ray emitter sources of fuel rods. In GET, maximum likelihood–expectation maximization (MLEM) was employed as an iterative reconstruction method. However, as convergence iteration in the algorithm is proportional to the pixel size, convergence is slow and the calculation cost for practical

An on-the-fly homogenization (OTFHom) method coupled with depletion based on Monte Carlo method was proposed and implemented in Reactor Monte Carlo code (RMC), which uses continuous energy cross sections to generate multi-group cross sections for specified problems and then uses them in Monte Carlo multi-group calculations to get more convergent results. In order to coupling with depletion, some modifications

Mapping of subassembly (SA) heads using down viewing transducers mounted on an under sodium ultrasonic scanner can be used for detection and evaluation of SA bowing. However, due to constraints in the rotating plugs on which the scanner is mounted, the down viewing transducers may not reach on top of all the SAs. For such a scenario, an ultrasonic glancing angle based imaging methodology (GAIM) is

In this paper, the deep learning based on the artificial neural network (ANN), which is referred to as the deep neural network (DNN), is adopted to build a new model for the generation of the resonance self-shielded cross-sections (XSs). In this model, using the dataset generated from the pin-based ultra-fine-group (UFG) calculations under a multi-dimensional parameter table, the multi-layer DNN is

The weight of spent nuclear fuel concrete storage casks usually exceeds 100 tons. Therefore, thermal testing of prototype casks is expensive and time consuming. In this study, thermal testing method using a scaled-down model of a storage cask was developed to perform the thermal testing of large-scale storage casks efficiently. Scaling analyses were performed to derive scale ratios between a prototype

Treatment of plutonium separated from spent fuel has been one of the most important issues in the utilization of nuclear power in Japan. This study investigates a molten salt accelerator-driven system (ADS) to transmute transuranic (TRU) nuclides to address the issue. This study proposes the MARDS (Molten salt AcceleratoR Driven System) concept which has four features: employment of chloride, no dedicated

The cyber security of NPPs recently has become a major issue, and a utility in Korea is requested to follow cyber security controls and to perform cyber security risk management based on the regulatory guide RS-015. Since it is difficult to fully implement the controls with limited resources, it is necessary to find which controls are relatively more important than other controls. The aim of this study

The China initiative Accelerator Driven subcritical System (CiADS) has already embarked on a plan to be built in south China’s Guangdong province for the research purpose of lead–bismuth cooled fast reactor and minor-actinide transmutation technology. In comparison with the other critical reactors, it is difficult to arrange the fuel handling equipment and to make the whole core refueling plan, since

Accurate and fast-running reduced order neutronics models are expected in many challenging applications such as multi-physics coupling analysis and data assimilations of reactors. The emphasis of the present study is to investigate an appropriate order reduction method for treating the movement of control rods. The empirical interpolation method is adopted to perform an affine decomposition to the

This paper investigates the application of the modified unscented transformation (MUT) method for uncertainty propagation in neutron induced cross section measurement in Nuclear Science. The method is the improvement on unscented transformation (UT) technique by incorporating more parameters (not random) to increase the number of sigma points. We examined the application of MUT method for covariance

After Fukushima Daiichi accident, the replacement of zirconium-based fuel cladding in Light Water Reactors (LWR) became one of the main challenges of the nuclear industry. Austenitic steel–clad presents some safety advantages comparing to zirconium alloys, noticeably, higher activation energy and lower enthalpy of metal-water reaction. Thus, it produces a slower hydrogen release into the containment

Thermal neutron scattering laws tabulated in File 7 for various materials the ENDF/B database require processing to generate the cross section and distribution data needed for use in Monte Carlo reactor simulations. Typically, the total scattering cross section is calculated on a fixed energy grid and interpolated on this grid. Nevertheless, this method may result in the loss of physical features in

The lead cooled fast reactor is considered as one of the most potential Generation IV nuclear systems due to the low working pressure, the appreciable neutron economy, and the considerable passive characteristics. Meanwhile, the lead bismuth eutectic (LBE) has the related advantages of lead with the weaker corrosiveness, which has been paid much attention by recent decades. Moreover, the sub-channel

This paper presents a series of methods designed to incorporate classical Probabilistic Risk Assessment (PRA) models such as Event Trees (ETs) and Fault Trees (FTs) into dynamic PRA. In contrast to classical PRA, dynamic PRA couples stochastic methods with system simulators to determine the risks associated with complex systems such as nuclear power plants. Compared with classical PRA methods, they

Soluble boron-free reactors inherently have larger reactivity feedback than conventional pressurized water reactors during power transients. The purpose of this study is to verify the capability of the long-term load follow operation with only control rods for the soluble boron-free small modular reactors. The effects of the long-term load follow operation on burnup distribution, power distribution

This study investigates for optimizing thermal treatment conditions for the effective decontamination of cobalt(Co)-contaminated concrete waste. About 21 g of ordinary Portland cement (OPC, Type I), 37 g of dry sand, 5 g of fly ash, and 24 g of aggregates were mixed, cured for 28 days, and used for characterization and thermal treatment. Thermal treatment was conducted at different temperatures between

In this paper, a sequential conjugate gradient method is presented to reconstruct the undetermined surface heat flux for nonlinear inverse heat conduction problem (IHCP). This modified method combines the merits of sequential function specification method (SFSM) and conjugate gradient method (CGM). The proposed method can effectively eliminate the leading error which is inevitably generated in SFSM

The lead-based fast reactor is very attractive as one of the Generation-IV nuclear energy systems due to high-power density, heavy liquid metal coolants with a high boiling point as well as harmless interaction with air or water. Nuclear heat is generated from very high neutron flux that is regulated and controlled by the drop of the Control Rods (CRs) for safe operation. The resisting effect of the

Flow-induced acoustic resonances in closed tandem side branches with large branch diameter are investigated experimentally. The diameters of the main pipe and branches are 305 mm and 64 mm, respectively. Three single branches with different lengths, L/d = 5.86–8.52, are investigated. The first and second hydrodynamic modes are excited at Strouhal numbers close to 0.4 and 0.8, respectively. The normalized

Numerical instability resulting from coupling isotopic depletion and neutronics is widely reported, but little work has been done to interrogate the phenomenon theoretically. This work extends the pioneering analysis developed by Densmore et al., who examined the stability of a monoenergetic neutron diffusion problem coupled with a simple depletion system through the explicit Euler scheme. Their Fourier

The micro heat pipe cooled reactor system with high efficiency energy conversion units are potential candidates for the energy support in decentralized markets, such as oceanic activities and remote regions. However, its development has been limited by several unfavorable factors. In this work, the potential opportunities and challenges of micro nuclear power system with high efficiency energy conversion

After a LOCA, debris enters the fuel assemblies through the sump strainers of the containment, increasing the pressure drop of the fuel assemblies. Experimental study of the effects of debris passing through the strainers and settling in the reactor core within the fuel elements on the core cooling is crucial. The distribution, attachment and blockage of debris in a fuel assembly are studied by the

The accuracy of the Eulerian two-fluid model to solve dispersed and segregated multiphase flows was evaluated. Five air-water tests representing flow regimes commonly found in nuclear installations were considered: (a) bubbles ascending in a stagnation water column, (b) air-water upward flow crossing an obstacle in a vertical column, (c) liquid-liquid co-current horizontal flow, (d) air-water counter-current

Nowadays, license renewal of nuclear power plants is difficult due to the political, technical, and economical reasons, and the optimum end-cycle strategy has become an interesting and important research topic. In this paper, commercial cycle optimization strategies considering permanent shutdown have been studied for a Westinghouse Electric Company type PWR core. Both the conventional equilibrium

Due to the high-dimensional integro-differential properties of the neutron transport equation (NTE) and the large-scale property of the nuclear reactor system, the detailed neutron transport simulation can be very time-consuming, which urges researchers to develop a fast and accurate technique. To improve this condition, this work develops a set of reduced-order models for the NTE (NTEROM) by combining

Best estimate plus uncertainty methodology is an advanced technology for licensing application of the nuclear power plants, but some aspects can still be improved in the conventional input-driven methodologies. For instance, uncertainties of the closure models are not properly handled; local sensitivity analysis methods not suitable for complex nuclear systems are still widely utilized; the sensitivity

The loose parts detection in low signal–noise-ratio (SNR) environments is investigated, which is frequently encountered in safety monitoring of the primary coolant system in nuclear power plant. The objective function method (OFM) based on high-order statistics is firstly applied to the impact response extraction under strong noise background. The parameter optimization of OFM is carried out to improve

In this research, the Operator-Splitting semi-implicit (OSSI) method, fixed-point implicit (FPI) method and Approximate Block Newton (ABN) method have been applied to the coupling of thermal-hydraulics code COBRA and neutronics code SKETCH-N. In the OSSI method, there’s only one data exchange within a time step. In the FPI coupling, an iteration convergence has been added to the OSSI coupling framework

Welding structures of primary coolant pipes were subjected on a series of accelerated thermal aging test at 400 °C until 30,000 h to investigate the resulting microstructure, impact and nanoindentation properties change. The results indicated that after long-term thermal aging, a mottled structure and precipitated G-phases were observed in ferrite phases both of straight pipes and welded joints, leading

Dynamic Monte Carlo (DMC) simulation of realistic nuclear reactors requires powerful variance reduction methods for even a few seconds of real time calculations. State-of-the-art numerical methods deal with the dynamic nature of the problem via successive Monte Carlo transport and TH (thermal-hydraulic) runs in a time step by time step manner. Such halting of the sample population at the beginning

The sodium-heated steam generator plays a crucial role in the operation and safety of sodium-cooled fast reactor. The configuration with multiple steam generators poses challenges to the feedwater control system design. It is necessary to adopt an appropriate control strategy. Based on the dynamic analysis, hybrid feedback and feedforward control theory is applied in the loop feedwater control. The

The JANUS Phase 7 shielding experiment performed in the NESTOR experimental reactor in 1991 studied deep fast neutron propagation in mild steel, boron carbide and sodium. Several steel plates, followed by boxes containing boron carbide and sodium were used as neutron shielding next to a fission plate. Gold, manganese, rhodium and sulphur detectors were positioned at several depths in order to measure

GSPN(0) is the lowest level approximation in the GSPN(K) theory, whose highest level GSPN(N) is equivalent to PN. GSPN(0) has the same differential equations as the traditional SPN but with different interface and boundary conditions. In this paper we show that GSP3(0) can be practically implemented and can indeed provide much improvement over SP3. The very popular traditional transverse integration

A fault diagnostic framework was investigated in this study for applications in thermal–hydraulic systems of nuclear power plants. The proposed framework consists of quantitative model-based diagnosis, statistical change detection and probabilistic reasoning. The use of physics-based diagnostic models provides high detection sensitivity and allows noise and measurement uncertainty to be incorporated

In a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactor (SFR), by supposing rather pessimistic condition, a large molten fuel pool would be possibly formed that contains enough fuel with the potential to exceed prompt criticality due to fuel compactive motion. Local fuel-coolant interaction (FCI) in the molten pool is identified as one of the typical inducing factors that may result in such

The heat pipe cooled reactor is an advanced reactor design which is nearly solid-state with an array of in-core heat pipes for passive heat transfer. The local heat pipe failure probability is certainly high over the reactor lifetime. Therefore, the reactor design must include consideration of heat pipe failure accidents which needs to be explored in-depth. The Monte Carlo neutron transport code RMC

Stability analyses of nuclear-coupled thermal hydraulics are carried out for SUPERSTAR – a natural circulation heavy metal (lead-cooled) fast reactor. Attributed to its natural circulation characteristic, flow stability as well as nuclear-coupled thermal hydraulics stability are of some concern. Modal expansion method as an extension for point reactor kinetics is used for neutronics and reduced order

Multi-group heterogeneous reaction rates calculated by the subgroup method in resonance calculations are not reproduced, when subgroup-collapsed one-group homogeneous calculations are conducted for the same problem, giving rise to a multi-group equivalence effect. In this paper, a new non-iterative equivalence method introducing partial current discontinuity factors (PCDFs) is proposed to resolve this

Safety and lightweight optimization are significant in developing high-performance compact nuclear reactors. We proposed the reactor radiation shielding optimization method by coupling genetic algorithm and neural network. The engineering design is a multi-objective and multi-constrained complex optimization problem. Thus, we improved it by non-dominated sorting genetic algorithm (NSGA-II) and mini-batch

This work applies the first order adjoint sensitivity analysis methodology to a reduced-order model of BWR dynamics to compute the exact first order sensitivities of the model’s state-functions with respect to the model’s initial conditions and parameters. These sensitivities are computed using exact forward and adjoint sensitivity functions, which are shown to yield identical results, within machine

The majority of this paper investigates the phase separation and pressure drop characteristics for a bubbly flow passing through a vertical branching T-junction. Effects of inlet superficial velocities of both phases and outlet pressure on the efficiency of phase separation in the T-junction are also considered. Based on the experimental results, the liquid separation efficiency increases as the inlet

Critical heat flux(CHF) is a significant parameter that determines the heat transfer capability of nuclear reactors, and therefore prediction of CHF with accuracy is of great importance for the design and safety analysis of nuclear power plants (NPPs). This paper presents a novel hybrid model based on Gaussian process regression (GPR) and ant colony optimization (ACO) for the prediction of CHF. In