The feasibility of the improved subgroup method coupled with the Sanchez-Pomraning Method (ISSP) is analyzed for the resonance self-shielding calculation of double heterogeneity (DH) problems in PWR. The Sanchez-Pomraning method is developed for both fixed source and the eigenvalue problems. The improved subgroup method adopts a fine-mesh energy structure to handle the resonance interference effect

The pellet bed reactor (PeBR) featured with high-power density, small reactor size and full retention of fission products, has promising applications for space nuclear power system. According to the special features of the PeBR for nuclear electric propulsion (NEP), a transient analysis code is developed, where the six groups delay neutrons point kinetics model, coolant flow and heat transfer model

Underwater gas bubbles are universal phenomenon in various industries. These bubbles appear under ocean conditions especially rolling conditions, that affect the mass transfer and produce different acoustic signals. The bubble formation and oscillation frequency characteristics under rolling conditions were investigated in this study by using a numerical method. Results showed that the bubbles followed

In the present work, the radioactive solution of 0.10 Molar 35S-aqua regia with half-life of 86.35 ± 0.17 days and specific activity of 519 Ci/cm3 was considered as a source. First, the beta spectrum of 35S was calculated. Then, by MCNPX code the effects of factors such as radius variation, height variation, type of substrate and type of side surface coating on the source efficiency were investigated

A methodology for determining fission source stationarity is presented that characterizes the behavior of higher order spatial moments of the fission source distribution using Functional Expansional Tallies (FET’s). The computation of these FET’s does not require a meshing scheme like traditional methods do (e.g Shannon entropy), and is computed natively within the Monte Carlo tallying routine. To

Nuclear fissile material was analyzed for the waste produced in the pyro-process. Nuclear waste is produced at each stage in the pyro-process, spent fuel head-end, major group recovery area and TRU-RE process. Each yields waste of different material composition and properties. Chemical analysis is basically determined for the measurement of nuclear material content. However, the possibility of non-destructive

Magnetite deposition on the secondary side of steam generators is a major concern in nuclear power plants. Therefore, it is necessary to mitigate the deposition using optimized water chemistry control. The purpose of this study is to investigate the effects of three pH control agents, ammonia, morpholine, and ethanolamine, on the magnetite deposition onto an Alloy 690TT tube under conditions similar

This paper presents the interaction between molten tin column and cooling water using the high-speed camera and the transient pressure measurement system. The effects of the molten tin column temperature, dropping height, diameter of molten tin column, and surrounding gas environment on the interaction of molten tin column with the water are examined. The results show that three modes of no explosion

In the corrugated plate separator, the droplets are separated by impacting on the wall of the corrugated plate, so it is necessary to study the droplet impacting on liquid film. In this paper, CLSVOF method is used to analyze the liquid droplets impacting on different thicknesses liquid film. The morphology of crown, the velocity field and pressure field distribution inside the liquid film have been

Passive safety systems are being considered in advanced reactor designs to provide inherent stability for the operation of the nuclear reactor. Passive shutdown is provided for guaranteed removal of decay heat under emergency conditions. The reliability of such systems should be very high. The static reliability assessment of such systems has been considered using various techniques such as fault tree

The higher eigen-modes of the forward and adjoint neutron diffusion equations are widely used in several application perspectives for engineering purpose, such as the stability analysis and generalized perturbation theory based sensitivity and uncertainty analysis. In this study, the higher eigen-modes of both the forward and adjoint neutron diffusion equations are calculated using IRAM based on domain

The Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark Progression problems #1 to #10 based on the 17 × 17 Pressurized Water Reactor (PWR) Watts Bar Unit 1 designed by Westinghouse have been proposed by the Consortium for Advanced Simulations of Light Water Reactors (CASL) as core physics benchmarks. These problems were selected to assist nuclear software and methods developers

It is well known that nuclear energy is a clean, safe, and reliable energy resource. Nuclear power plant is one of the prominent applications of nuclear energy. In this paper, aiming at improving the control performance for power level of nuclear power plant, an extended state observer-based adaptive dynamic sliding mode control (ESO-ADSMC) scheme is proposed. The mathematical model of nuclear reactor

Flow boiling heat transfer under marine motions is an important research area that provides more opportunities for the development of offshore applications. In this paper, a comprehensive review of the literature on flow boiling under marine motion is presented, including experimental and numerical studies. The effects of marine motions on heat transfer coefficient, critical heat flux, pressure drop

In this paper, a framework of improved risk monitor system that integrates multiple applications of retrospective investigation analysis of risk factors, point-in-time risk monitoring and proactive risk management is presented. Among these applications capabilities, the plant configuration risk management is implemented and demonstrated with a safety injection system in the event of medium-break LOCA

In this study, the nuclear-data adjustment method based on the continue-energy cross sections has been implemented for the fast reactor. The technique of nuclear-data adjustment based on the sensitivity/uncertainty analysis has been widely applied for the fast reactor to improve the nuclear-data accuracy. The previous researches towards nuclear-data adjustment were implemented to the multigroup cross-section

This paper presents a living PSA modeling and updating method based on a time-dependent modular fault tree (MFT), which takes into account all the failure modes involved in the whole life cycle of a component for continuous state transition and time-dependent problem analysis. And a Stage 4-Living PSA, characterized by real-time online automatic updating model, is achieved using a combination of the

Dynamical behavior of the Brazilian Multipurpose Reactor (RMB) is investigated by using the coupled neutronics and thermal hydraulics code NTHC1, which consists of the neutron point kinetics (NPK) equations with six groups of delayed neutrons, the heat conduction equations in a fuel plate, and the energy transport equation in a coolant subchannel. The Serpent 2 Monte Carlo code is employed to calculate

Nuclear graphite is widely used in traditional and advanced nuclear reactors. Especially in High Temperature Gas-cooled Reactors (HTGRs), nuclear graphite is the main composition of fuel element and reflector. The graphite cross section data libraries used in practical reactor physics calculations are generally evaluated from ideal single crystal graphite material. However, the actual nuclear graphite

In nuclear fuel assembly mixing vanes are fabricated on the spacer grids to promote diversion cross flow and turbulent mixing. In order to investigate the decay of diversion cross flow and turbulent mixing induced by the split mixing vanes, particle image velocimetry (PIV) measurement is carried out to acquire high-fidelity flow field data in a 5 × 5 rod bundle with mixing vane spacer grid. Intensities

The reactor fuel consumption calculation is an important part of reactor design and analysis, which usually needs to be completed iteratively by transport calculations and point-depletion calculations. AMAC is a newly versatile point-depletion and radioactive-decay code for use in simulation nuclear fuel cycles and calculating the nuclide compositions and radiation characteristics of materials. The

In this work the mathematical derivation and numerical analysis of a fractional nuclear reactor point kinetics in time and space (TSFNPK) is presented. The TSFNPK model was derived considering a non-Fickian law for the neutron density current where the differential operators in space and time are of fractional order. The TSFNPK equations presented in this work constitutes a novel model for nuclear

Hydraulic load analysis is of great importance in the design of pressurized water reactor. It provides load conditions for structural mechanics analysis of primary loop system and ensures the reliability of primary loop pipeline and main equipment. Currently common hydraulic load analysis software mainly includes RELAP5 and WAHA, both of which adopt the classical two-fluid six-equation single pressure

In this paper, first, MELCOR calculations for a conventional high power pressurized water reactor are performed to obtain the geometrical configurations and remaining fraction of the fission products in the degraded reactor core during the early phase of a severe accident. Then, based on the coupling between MELCOR and Serpent 2 code, sensitivity analyses on the criticality are performed for various

This paper investigated the Hopf bifurcation and post-instability of tube bundles subjected to two-phase cross-flow and loose support. A nonlinear dynamic model of a single tube in tube bundles considering the effects of two-phase flow force and impact force between the tube and tube support plate was established. Based on this model, a numerical simulation program was written to calculate the bifurcation

Uranium mononitride (UN) is a nuclear fuel material of interest in the design of advanced reactors. Phonon spectra and dispersion relations of UN in its anti-ferromagnetic and paramagnetic structures were calculated using ab initio lattice dynamics. Subsequently, the dynamic structure factors of uranium and nitrogen in UN were generated in the incoherent approximation using the phonon expansion method

We study effects on temperature coefficient with three different minor actinide (MA) loading patterns on PWR burnable poison rods. MA transmutation on PWR burnable poison rods ameliorates FTC and MTC. Three different MA loading patterns on PWR burnable poison rods cause no distinct difference to FTC and MTC in normal operation temperature range. This indicates that mixing MA with burnable poison homogeneously

The planar Method of Characteristic (MOC) solution based direct whole core calculation method is stabilized and extended by employing an axial one-dimensional (1D) MOC solver with several augmentations. In order to resolve the inherent instabilities involved in the coarse mesh finite difference formulation employing a nodal simplified P3 (SP3) axial solver, the factors causing negative fluxes are identified

Experimental investigations on single-phase convection and critical heat flux in a vertical tube with an inner diameter of 6.0 mm under inlet flow oscillation conditions were carried out for the following range: 255 ≤ Gav ≤ 455 kg·m−2·s−1, 0.5 ≤ P ≤ 3.0 MPa, 0 < ΔG/Gav ≤ 3.0 and 0 < τ (period of flow oscillation) ≤ 10.6 s. Single-phase convective results show that the fluctuation amplitude of single-phase

Accurate prediction of rod-bundle critical heat flux (CHF) is a main challenge in thermal-hydraulic design and safety assessment of Pressurized Water Reactor (PWR). The most widely applied approach is empirical correlation developed based on rod-bundle experiments carefully tested with geometry and power distribution representative to PWR working conditions. The standard procedure is to correlate the

The melt pool could form in the nuclear reactor lower plenum under a hypothetic severe accident. In such condition, remaining the integrity of reactor pressure vessel (RPV) is treated as one of the severe accident management strategies. The investigation of melt pool natural convection behavior is one of the most important aspects in the in-vessel retention (IVR) investigation. Various experimental

In nuclear power plants, core melt accidents are often accompanied with fuel–coolant interactions (FCI), which may escalate to steam explosion destroying the integrity of structural components and even the containment under certain conditions. The vapor film and two-phase region formed by boiling around the molten jet at high temperature have a great influence on the fragmentation of the jet and the

This work shows basic features of pebble flow in HTR-PM via DEM simulation, in accompany with pebbled beds of varied base angles, friction coefficients and recirculation modes. The effects of base angles, friction coefficients and recirculation modes on the characteristics of pebble flows are explored. Beside phenomenological observation by pebble stripes, the three-dimensional pebble spindles and

The wire-mesh sensor (WMS) is a kind of advanced instrument for gas-liquid two-phase flow measurements. In previous research, the sensitive volume was assumed to be uniformly distributed in the cross-section. The received current was assumed to represent the conductivity averaged over a uniform sensitive volume (USV), which is defined by symmetry considerations within the geometry of the electrode

Verification and validation of multi-physics codes dedicated to fast-spectrum molten salt reactors (MSR) is a very challenging task. Existing benchmarks are meant for single-physics codes, while experimental data for validation are absent. This is concerning, given the importance numerical simulations have in the development of fast MSR designs. Here, we propose the use of a coupled numerical benchmark

Recent advances in nuclear fuel materials research, particularly on the topic of accident-tolerant fuels, have brought up potential opportunities for expanding the operating envelope of existing light water reactors. As many of the performance improvements offered by these technologies may be most fully realized by increasing fuel enrichment beyond the standard 5% limit, this paper examines the potential

A phenomenological model to account for the effect of a grid spacer with mixing vanes on CHF has been developed. It takes into the account of the mixing vane configuration (i.e., angle, number, position, etc.) on the flow field, the turbulent mixing between subchannels and heat transfer. Changes in flow field, turbulent mixing and heat transfer have led to CHF enhancement in subchannels and hence the

We present effective periodic surveillance test (PST) strategies enhancing the availability of nuclear safety systems with three-channel configuration in PWR (Pressurized Water Reactor), which can be used in limited space applications. The suggested PST strategies are about testing the system with the optimal surveillance test interval, the different way depending on the system operation time and testing

Uncertainty and sensitivity analyses are required in fuel cycle analyses for studying the viability of electronuclear scenarios by means of their response to variations in the input parameters. This implies that nuclear fuel cycle simulators have to be extended for dealing with this kind of problems in order to become reliable tools. However, given their complexity, the excessive computational effort

To figure out the flow characteristics through cracks with narrow and short passage, experimental and theoretical studies were conducted in this study. The leakage flow phenomena such as crack opening enlargement, partial flashing and actual flow region were discovered during the test. The mechanism models were proposed according to the test results. Mechanical model on the COD (crack opening displacement)

Digital instrumentation and control (I&C) systems have been replacing analog I&C systems in nuclear power plants (NPPs) on account of the beneficial functions digital systems provide. As a result, NPP risks have been centralized into the digital I&C systems, with the issue of cyber-attack thus emerging as one of the new threats. To achieve more effective cyber security, it is necessary to evaluate

Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator

In this paper, a Roe-type Riemann solver is developed for the two-fluid seven-equation model. The general form of equations of state (EOS) is under consideration in order to simulate water-steam flow. To avoid the difficulty of mathematical treatment, the governing equations system is transformed into a equations system in which the primary variables are density, velocity, entropy and phase volume

Diamond crystal-based detector for neutron flux measurement is a promising solution as a fast, small and stable detector especially suited for fast transient measurements. The work aims to test the method and to find the best parameters of the pulse shape analysis for the thermal neutron detection. The developed diamond detectors were tested with the AmBe neutron source and in the VR-1 reactor core

The Zirconium isotopes have been widely used in many PWR fuel cladding materials. Ignoring the resonance self-shielding effect for Zirconium isotopes, it may introduce 100–300 pcm errors in PWR reactivity predictions. The reference dilution cross section, equivalence theory, and subgroup method can be used in treating resonance self-shielding effect for Zirconium isotopes. Hence, it is important to

This paper demonstrates the potential use of a method of power profile determination using Cerenkov radiation to detect and identify some physical anomalies in a reactor. Specifically, the blockage of a coolant channel in a square lattice TRIGA reactor is investigated. The ability to quickly detect some blockages is demonstrated. A method is developed through which the blocked channel and approximate

The purpose of this study is to extend a method of power profile reconstruction based upon Cerenkov radiation measurements outside of a reactor core to provide information about the 3D power profile. Multiple simulated viewpoints above a coolant channel in a TRIGA reactor are used to estimate the axial tilt in the power density in the surrounding pins. The ability to produce a map of the axial flux

Prediction and analysis of coolant flow and mixing problems in the reactor pressure vessel (RPV) are some of the most critical studies to ensure system safety and balance of nuclear power plant. The flow mixing problem in the reactor primary system with complicated interior structure is a large domain three-dimensional problem. In this paper, we investigated the performance of utilizing multiple RANS

Fission fragments emitted in a fissile solution create tiny gas bubbles, the size of which is determined by the linear energy transfer (LET) of the particles. The LET of fission fragments of 235U in aqueous solutions of uranyl nitrate has been determined, and using methods adapted from the literature, the size of gas bubbles generated along the tracks of these particles has been estimated, revealing

In most designs, the coolant channel is under non-uniform heating, mainly one-sided heating. There is no clear understanding of the problem of unilateral heating. In order to deepen the understanding of the mechanism of CHF in the unilateral heating rectangular narrow channel, a three-fluid transient model of annular flow was developed for the rectangular narrow channel heated on one side. Basic conservation

Escape probabilities are fundamental inputs in the calculation of self-shielded cross sections for multigroup neutron transport models. Rational approximations are often used to compute escape probabilities in geometries commonly encountered in nuclear reactors. Although analytical approximations are simple and rapid, they are limited in applicability and accuracy. Here we show that ray tracing can

To increase the surface wettability is a feasible way to enhance critical heat flux (CHF) according to some CHF theories and many previous researches. It is found that surface wettability of metal can be increased after the surface being irradiated by Gamma-ray irradiation and CHF in pool boiling can thus be enhanced. In this study, whether CHF in downward-facing flow boiling can be enhanced after

This study presents a numerical assessment of the effects of tensioning and curvature on the impact resistance performance of a prestressed concrete wall. Specifically, a prestressing method and numerical analysis method applying an impact load-time function are proposed and described. To transfer of initial prestress to the wall, the prestressing method employs a thermal load and appropriate boundary

The sensitivity of FCM parameters to the equivalent radius of the RPT model is evaluated by a quantitative method, and six important FCM parameters are identified. Then the relationship between the six parameters and the equivalent radius of the RPT model is analyzed. Finally, a rapid fitting RPT method is proposed. The new method can directly establish the RPT model by a fitting formula without relying

Load rejection is an extreme transient that may lead to a reactor scram and further threaten the safety of Nuclear Power Plants (NPPs) and thus should be well investigated. This paper presents the full load rejection transient simulation, the worst condition of load rejection transients, on the CPR1000 NPP. A whole plant simulation platform was first established. Then load rejection transient from

Proper generalized decomposition (PGD) can be described as a numerical extension of separation of variables and thus be employed as a solution technique for multi-dimensional problems, where dimensions should be understood in the phase-space context of the governing law at hand. This paper presents a PGD approach on efficiently solving problems involving multigroup neutron diffusion. Two PGD approaches

In recent years, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation

The optimization of an emergency operating strategy necessitates the risk quantification of mitigation action. However, there are challenges associated with plant risk estimation: 1) there are thousands of accident scenarios in given conditions that will cause core damage, and 2) the time-dependent recovery probability and various available times of component recovery according to plant conditions

Gas-liquid jet flow is a typical phenomenon in severe reactor accidents. It is great significance to evaluate the scrubbing efficiency of aerosols in jet flow conditions reasonably for the analysis of source terms and also adopting mitigation measures. In this paper, a visualization experiment is carried out, and a mathematical model is developed based on the experiment results by considering the deposition

Mesh distribution is an important part of the calculation in many thermal hydraulic system codes. Changing the mesh size is a common way to improve the calculation accuracy and efficiency. Automatic adjustment of the mesh size is an essential feature of adaptive mesh methods, among which the moving mesh method is a typical one. With the moving mesh method, mesh number can stay the same when adjusting