In this study, a thermal-hydraulic analysis code has been developed for primary system of the integrated small modular pressurized water reactor IRIS (International Reactor Innovative and Secure). Numerical model of the primary system has been established, and characteristics of equipment and layout are considered. The sub-channel model is used for performance analysis of hot channel in the core. In

An uncertainty analysis based on the Latin Hypercube Sampling (LHS) method and MELCOR code for hydrogen production during a SA in WWER1000 nuclear reactor is presented. Firstly, the total amount of the produced hydrogen during the SBO with Large Break Loss of Coolant Accident (SBO-LBLOCA), SBO with Small Break Loss of Coolant Accident (SBO-SBLOCA) and Station Black Out (SBO) is estimated by the MELCOR

In this paper, the elliptical drop deformation process in ambient liquid under an impact was systematically studied based on an experimental method combined with computational fluid dynamics (CFD). The effects of nondimensional parameters (i.e., shape parameter e, viscosity ratio η, and density ratio ε) on the vortex dynamics during the drop deformation process in ambient liquids were systematically

A neutron counting (NC) technique for in-beam active neutron interrogation has been developed and deployed at the National Research Universal (NRU) reactor of the Canadian Nuclear Laboratories (CNL). A distinguishing feature of the method is that the sample to be interrogated remains stationary during the irradiation and counting periods for the detection of prompt and delayed neutrons. The NC system

Up to now, different methods have been developed for estimation of buildup factor (BF). However, either expensive estimation or time-consuming estimation are major restrictions/challenges of these methods. In this study a new technique utilizing combination of Monte Carlo method and the Bayesian regularization (BR) learning algorithm of multilayer feed-forward neural network (FFNN) is employed for

The aim of this study is to analyze the effect of radial void fraction distribution, considered at three degrees of fidelity, on lattice k∞ and 2-group homogenized cross sections. The novelty of this work lies in its use of experimentally-determined radial void fraction distributions which allows for fine-scale coolant density data with no predictive uncertainty and which could be prohibitively expensive

A 3-D full reactor core simulation of VVER-1000/V-446 during full power normal operation is developed to investigate reactivity control and heat transfer improvements when using silver nanoparticles as the base coolant. Coupled neutronic/thermohydraulics analysis is performed using efficient coupling method in which the detailed neutronic model has been created in MCNP, and porous medium method has

The direct whole core calculation is applied on Cobalt-59 activation calculation. The coupled neutronics, thermal hydraulics and depletion calculation is conducted, where the neutronics is solved by the Monte Carlo code JMCT and the thermal hydraulics is treated with sub-channel method. In the neutronics calculation, the reactor core is pin-by-pin modeled and the cobalt rods are modeled without approximation

The treatment of chemistry and thermodynamics in the integral severe accident codes is typically limited. A more accurate treatment of the chemistry during the severe accident modelling is, therefore, of great interest. For this purpose, the work is focused on the development of coupling of the MELCOR code with chemical thermodynamic calculations using GEMS codes and HERACLES database. Developed coupling

In this study, the application of nuclear transmission based techniques to the measurement of two-phase flow was evaluated and the experiment setup was described. Because neutrons are appreciably affected by interaction with hydrogen atoms, techniques based on transmission of thermal and fast neutron are particularly suitable for use with water or any other hydrogenous fluids. Based on the effective

Molten salt reactor (MSR) has good economic potential, inherent safety and nonproliferation characteristics. In this paper, the 450 MWth graphite molten salt reactor is taken as the research object. First, the neutronic design of a 450 MWth graphite molten salt reactor is introduced in terms of the core material, conceptual design, static physical characteristics and core design parameters. Then, the

Subcritical noise experiments at frequencies up to 100 kHz employing two distinct boron dilution in the reactor tank water (286.8 and 578.6 ppm of natural boron) were performed in the IPEN/MB-01 reactor. The Auto Power Spectral Densities (APSD) were inferred employing the IPEN/MB-01 Correlator and were best described by a four-mode decay model up to about 70 kHz. The analyses reveal that the first

Canadian Nuclear Laboratories (CNL) has many decades of thoria research and development experience, including fabrication, irradiation testing, and post–irradiation examination. CNL has previously shown fission gas release (FGR) to be dependent on fuel microstructure in thoria fuels. Lower density fuel with large granules surrounded by porous regions results in higher FGR. Higher and more uniform density

The present study provides recommendations for choosing the appropriate moving mesh techniques of CFD to describe the dynamic behavior of a check valve and to evaluate a pressure build-up in response to the valve’s closing behavior. Four different moving mesh techniques of CFD were modeled and compared, and the compressibility of the fluid was involved in simulating the pressure build-up by valve closing

Based upon relevant statistics, it was suggested that potential scenarios for uncontrolled recycling of disposed activated metal waste, once it is excavated out of a near-surface repository, would not be excluded without prudent deliberation. Recycling a by-product slag from a smelter in parking lot foundation shows the highest radiation dose per unit activity concentration for 94Nb among combinations

The scenario code CLASS relies on infinite assembly simulation to predict fuel actinide inventories at exit burnup. In the current work, we replace these assembly calculations by full-core simulations and evaluate the impact on actinide inventories predicted by CLASS. To achieve this goal, we generate neural network training databanks for CLASS using the lattice code DRAGON5. For UOX fuels, the databanks

Validation of condensation model of CFD code FLUIDYN-MP has been carried out against two experimental data. The influence of turbulence model and grid size on condensation rate has been studied. Models for the simulation of wall condensation in presence of non-condensable have been developed to simulate the typical post-LOCA steam laden atmosphere of the containment of a Nuclear Power Plant (NPP).

Pressurized Thermal Shock (PTS) in the Reactor Pressure Vessel (RPV) can occurs during injection of Emergency Core Cooling (ECC) water into the cold leg of Pressurized Water Reactor. Injected ECC water generates large temperature gradient which leads to large thermal stresses in the RPV wall. To study this behaviour in Rossendorf Coolant Mixing Model (ROCOM), case of buoyancy driven mixing with a constant

The Modular Accident Analysis Program version 5(MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of AP1000 with the Modular Accident Analysis Program version 5(MAAP5). The pressurizer (PZR) safety valve stuck-open is essentially a small break loss-of-coolant accident which

A depletion solver has been implemented in OpenMC and is described herein. The depletion solver is implemented in Python and interfaces with OpenMC’s transport solver through a C++ application programming interface, which enables an in-memory transport-depletion coupling. Multiple integration methods for advancing in time have been implemented and exhibit tradeoffs in cost, accuracy, and memory use

The model of the EVEMP (Extra Vessel Electro Magnetic Pump) with a mass flow of 5000 kg/s and a developed pressure of 15000 Pa in the 350 °C operating temperature has many advantages as a coolant circulation pump for the Micro URANUS. Micro URANUS is an SMR with a rated output of 60 MWt which is abbreviated as ‘Micro Ubiquitous, Rugged, Accident-forgiving, Nonproliferating, and Ultra-lasting Sustainer’

Variance reduction technique is one of the most efficient methods to accelerate Monte Carlo (MC) results convergence. Among all techniques, MC self-couple variance reduction method is highly praised for it does not need any secondary model or code. This method can be applied in high-fidelity geometry simulation. In this paper, a new mesh-based bias parameters generation method based on multi-group

Depletion solver with Chebyshev rational approximation method (DESCAR) was developed by Nuclear Power Institute of China (NPIC) to solve depletion equation. In this paper, the algorithms for solving depletion sparse matrix are studied, including symbolic Lower-Triangle and Upper-Triangle (LU) decomposition using elimination directed acyclic graphs (EDAGS), symbolic LU decomposition using symmetric

Visualization experiments and mechanism analysis were performed to fully investigate the thermal hydraulic behaviors during bottom reflooding within narrow rectangular channel. Qualitative and quantitative analysis of the two-phase flow regimes were conducted by means of high-speed camera. Conclusions can be obtained that five flow regimes were observed, including dispersed flow, inverse annular flow

The new small modular reactor (SMR) NuScale has been in the approval process for Nuclear Regulatory Commission by its off-site fabrication and assembly and good safety performance. To ensure the safety of the fuel rod the rod ejection accident at the beginning of cycle (BOC) and the end of cycle (EOC), the traditional fuel rod is replaced by accident tolerant fuels (ATFs). Neutron physics and thermal

There is limited validation of MONK using 232Th/233U benchmarks, and existing validation of Monte Carlo codes has generally used older nuclear data libraries. Here we use MONK-9A to systematically study a series of thermal 232Th/233U-based benchmark experiments, taken from the International Criticality Safety Benchmark Evaluation Project handbook and using JEF-2.2 and JEFF-3.1 libraries. We find that

UF6 gas flow in the total regions of a rotor with the radius of 0.1 m and the length of 1 m in axisymmetric and steady states was simulated using a new hybrid CFD-DSMC method in the OpenFOAM framework. Implicit coupled density-based scheme was performed for CFD method, and Variable Hard Sphere (VHS) and diffuse model were employed in DSMC method. Also, as an initial estimation, the local Knudsen number

Accident tolerant fuels (ATFs) are designed to mitigate damaging interaction of hot steam and fuel clad during off-normal condition. These materials can withstand against high temperatures for longer period of time. They have lower oxidation kinetics, higher resistance to core degradation and lower hydrogen pick-up. The main purpose of this study is thermal-hydraulics evaluation of ATFs during an unmitigated

This paper paves the way to numerically solve the k-eigenvalue neutron transport equation with complex variable arguments, and then employs the developed complex variable transport solver to calculate the k-eigenvalue sensitivities with respect to nuclear cross sections using the complex-step derivative method (CDM). CDM utilizes the Taylor series expansion in the complex plane whereby the imaginary

The successful decommissioning of nuclear facilities requires the derivation of DCGL in accordance with national site release criteria. MARSSIM, a widely used decommissioning guideline, recommends that a probabilistic analysis should be performed to derive the DCGL, which can be done by using the RESRAD code. Among the RESRAD codes, the RESRAD-OFFSITE code is the latest software edition and can derive

In this paper, we collected the analysis data of radionuclide concentrations and performed verification analysis on the data integrity and outliers in radioactive waste. These analysis results investigated the correlation between the radionuclide concentration data. As a result of analyzing the characteristics and outliers of the radioactive waste radionuclide data by using the collected original and

The transient process of the gravity-driven draining in the core make-up tank (CMT) system will be affected by the wall-stored energy, which will lead to the deviation of experimental results based on scaled-down test facilities. In order to understand the corresponding mechanism, the Hierarchical Two-Tiered Scaling (H2TS) method was adopted to scaling effects. The Relap5 code was used to simulate

In an attempt to explore the significance of transport theory in neutron noise, localization of a noise source by Green’s function based on transport theory is investigated. There are considerable differences between Green’s functions based on diffusion and transport, such as small dimensions, near edges, high heterogeneity medium and high-frequency source of perturbation. These differences are expected

The natural circulation loops have been widely used in the design of next generation nuclear reactors and spent fuel storage facilities. However, the friction correlations for fuel bundle and grid spacers in RELAP5 code are not applicable for natural circulation flow. This paper summarizes the modeling methods in previous researches, and proposed a flow resistance model for rod bundle and grid spacers

The station blackout accident is clearly a very serious problem that jeopardizes the safety barrier of nuclear reactor and ascends the core integrity steeply. The dual-cooled annular fuel has been introduced to boost the generated power density and safety margins. The fuel is cooled both internally and externally by allowing the liquid to flow on inner and outer heated surfaces. The fuel peak temperature

Human reliability analysis (HRA) is a systematic technique to assess human systematic risk, and has been widely used in various industries for enhancing the safety and reliability of complex socio-technical systems. However, there is a lack of a comprehensive bibliometric analysis of the literature on this topic. This article aims to provide academics and practitioners with a bibliometric overview

Flow blockage in fuel assemblies is a severe accident that must be considered in reactor design and safety analysis. Porous type internal blockages are more likely to occur within fuel assemblies by the accumulation of suspended impurities following coolant flow. For lead-based reactor, the deposition of oxide impurities with low thermal conductivity are expected to represent a safety concern. In this

A sub-channel code capable of modelling the flow blockage of the annular fuel has been developed and verified by the experiment. A fuel conduction model has been developed for the blockage to consider the effect of the blockage on the temperature distribution of the fuel. Blockage at different position has been modeled and the middle blockage is selected as the calculation object. Blockage of inner

The unique feature of molten salt reactor (MSR) is that heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture which acts as both fuel and coolant. The fission energy from fissile fuel is released immediately into the fuel salt in the core, and the delayed neutron precursors (DNP) have a loss due to fuel circulation in the primary loop, which both

Breakdown voltages of helium between needle-plane copper electrodes were measured to obtain the base data for insulation design of the electrical equipment working in high temperature and pressure helium gas in the High Temperature Gas-cooled Reactor (HTGR). The measurements were performed at a gap length between 1.02 mm and 5.30 mm, temperature and pressure varying from 0.01 to 7 MPa and 25–180 °C

The proposed Canadian SCWR is based on the concept of the CANDU reactor. However, the dynamic characteristics of the SCWR significantly differ from the CANDU reactor because of the special heat transfer phenomenon of the supercritical water in the fuel bundle. Thus, it is important to design a control system to maintain the SCWR operating at the required operating point when it is subjected to disturbances

An evolutionary simulated annealing (ESA) method has been developed for the problem of fuel loading optimization of VVER-1000 reactor. The ESA method improves original simulated annealing by using crossover and mutation operators to generate new trial loading patterns (LPs). A core physics calculation code for fuel LP optimization of VVER reactors (LPO-V) has been developed and verified based on a

Selection of a perfect nuclear reactor fuel demands the neutronic analysis of the available fuel materials in particular reactor environment. This paper summarizes the pin-cell analysis of eight (08) models considering five (05) different chemical compositions of nuclear fuel, three (03) different neutron moderator and two (02) types of cladding materials using the open-source Monte-Carlo code OpenMC

The ultimate purpose of the nuclear power plant safety is to protect the public and the environment from unacceptable radiation exposure in all operational states. This paper presents a novel severe accident management framework to cope with the risk of radioactive material release to the environment. The objective is to reduce the contribution of severe accidents to the large and early release frequencies

Responding to the rapid growth of size reduction in engineering systems, it is crucial to explore the safety performance of supercritical fluid systems in miniature tubes. In this paper, numerical studies are conducted to explore the performance of baffles on heat transfer deterioration (HTD) mitigations for supercritical CO2 (sCO2) flowing in a horizontal mini tube using RNG k-e turbulence model with

A new particle-transport code NECP-MCX has been developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi’an Jiaotong University. NECP-MCX is aimed at the simulation of deep-penetration problems, including reactor radiation-shielding calculation, fusion reactor blanket calculation, etc. These problems are challenging for the Monte Carlo (MC) method, which needs a huge number of

Key parameters of Pebble Bed Advanced High Temperature Reactor (PB-AHTR) core are studied, and then the preferred core conceptual design is proposed. Both evacuating molten salt and injecting poison into coolant can be used as the auxiliary system of the second shutdown system in PB-AHTR. Compared with injecting poison into coolant, the effect of evacuating molten salt on the core is smaller and more

The determination of non-fundamental modes of the diffusion equation is required for computing CANDU reactor power distribution from analysis of in-core detector readings. They are also important for understanding subcritical mode instabilities occurring in boiling water reactors. The legacy method for computing these modes is the Hotelling deflation technique based on bi-harmonic decontamination.

It is a well-known fact that the physical processes taking place in nuclear reactors currently operating are addressed based on the Maxwell–Boltzmann distribution. However, for new reactor designs that can be developed, as well as for any operational condition beyond the basis of the project, it may be relevant to consider another type of velocity distribution. In recent papers, for the evaluation

In this study, a mathematical model has been developed for oxidation of pellets of uranium dioxide in the air at the bulk temperature range between 605 and 727 K. Oxidation of this pellet causes an increment in the size of the UO2 particles, decrement of the pellet porosity and breaking-down of this pellet, subsequently. Therefore, the presented model has two stages for this oxidation process, before

Computer tomography (CT) is a useful probe to measure two-phase flow parameters. Because of the dynamic characteristic of the two-phase flow, the high temporal CT is required. A compact CT system, consists of seven isotope γ-ray sources, seven groups of cadmium zinc telluride (CdZnTe) detectors and collimators, is proposed. The 241Am is selected as the source by balancing its ability to distinguish

The reactivity of a fuel assembly including burnable absorbers can become the largest at low fuel burnup, so the accuracy of reactivity calculations for such systems is important. To investigate this issue, the impact of gadolinium isotopes’ nuclear data on neutron multiplication factor k is quantified. Sensitivities of k to nuclear data are calculated from 0 to 20 GWD/t for a BWR 3 × 3 multicell model

Automatic differentiation (AD) is a set of techniques which allows the numeric evaluation of derivatives of functions calculated by a computer program. In recent years, interest in AD has grown significantly in many disciplines, especially in the context of gradient-based optimization algorithms. Sensitivity analysis is another natural application area for AD methods. However, despite the large body

We develop an analytical model about concrete ablation during ex-vessel cooling in a pre-flooded cavity. The model considers the melting of dry debris bed, the formation of crust, the growth of crust and the thermal ablation of concrete. This model can obtain concrete ablation depth during ex-vessel cooling. It is evaluated the adequacy of ex-vessel cooling in the pre-flooded reactor cavity of Westinghouse

The Continuous Galerkin Virtual Element Method (CG-VEM) is a recent innovation in spatial discretisation methods that can solve partial differential equations (PDEs) using polygonal (2D) and polyhedral (3D) meshes. This paper presents the first application of VEM to the field of nuclear reactor physics, specifically to the steady-state, multigroup, neutron diffusion equation (NDE). In this paper the

The scope of this work comprises the study of the mathematical nature of the RELAP5 two-fluid model (TFM) for two-component (water and non-condensable gas (helium)) and two phases (liquid and vapor) mixture flows, and to predict the accuracy of the model under these flow conditions. The method of characteristics is applied to check the non-hyperbolic nature of conservation equations of RELAP5 for some

A predictive approach to fault diagnosis in complex systems such as the Nuclear power plant (NPP) is becoming popular because of the efficiency and accuracy it presents. However, there is still a huge gap between the proposed fault diagnosis techniques and engineering applications. To further optimize the fault diagnosis route and encourage real-time application, this paper presents a highly accurate

As nuclear power plants tend to be miniaturized and integrated, equal-height-difference natural circulation systems (EHDNCSs) have potential applications in floating nuclear power plants. This work experimentally investigated the flow modes of an EHDNCS at low pressure. It has observed four quintessential flow modes in the experiments. The transient characteristics of each flow mode were analyzed by

To date, studies on nuclear emergency evaluation focus on emergency preparedness and medical rescue ability. The evaluation of Rescue Measures at containment vessel is critical for Nuclear Power Plant (NPP) emergency response under the severe accident. In this paper, a comprehensive evaluation index system was established for rescue measures of China national nuclear emergency engineering rescue team

Improved computer capabilities and parallel algorithms have enabled high fidelity coupled analyses of classical nuclear power plant designs, which can also be applied to analyze the new heat pipe cooled reactor design. The neutronic/thermal coupling is a key part of multi-physics, high-fidelity reactor simulations. In a heat pipe cooled reactor, the solid-state core can reach temperatures over 1000 K