Traditional reactor system analysis tools are confronted with challenging difficulties in model development, low accuracy, and poor convergence. To solve these problems, the automatic differentiation (AD) method that allows the automatic numerical calculation of derivatives of functions was adopted to develop the reactor system code in this paper. For the simulation of single-phase models, the steady

Transition from conventional analog technology to advanced digital technology is a global trend, and commercial and research reactors are increasingly using digital instrumentation and control (I&C) systems. However, common cause failures (CCFs) of the digital platform are significant issues. In this paper, we proposed a potential solution for such CCFs by implementing a digital hybrid model that combines

In the online fission gas removal system applied to TMSR, the venturi bubble generator was developed to produce the dispersed bubbles from the reactor coolant. It is essential to establish the critical condition of bubble breakup for the bubble generation. Numerical simulations of the bubble dynamics in an unsteady turbulent flow in the bubble generator are performed with Volume-of-fluid (VOF) and

A short-term station blackout (STSBO) accident has been considered as one of risk dominant sequences for nuclear power plants (NPPs). Unless appropriate mitigation actions are not taken, a rapid damage of the reactor core and thereby a severe accident may result in. This paper presents best practice analysis results on the evolution of the severe accident sequences and fission product (FP) source terms

As the share of renewable energy will be increasing, there is a growing interest in flexible power sources and energy storage systems due to the intermittent nature of renewable energy. We propose the Energy Storage System (ESS) integrated Nuclear Power Plant (NPP) as a solution. The system operates on load-following by branching the mass flow from the steam cycle to the ESS without changing the reactor

The corrosion results in not only material degradation but also causes CRUD generation in the primary coolant. The axial offset anomaly (AOA) or CRUD Induce Power Shift (CIPS) is caused by CRUD accumulation on the fuel assemblies. It also causes an increase in Occupational Radiation Exposure (ORE). Therefore, in this work, activated CRUD behavior due to 54Mn, 59Fe, 58Co, 60Co, and 51Cr has been studied

Transuranic (TRU) elements can be used as an alluring fuel in a graphite-moderated Molten Salt Reactor (MSR) with liquid fuel and online reprocessing. However, the positive temperature feedback coefficient (TFC) stemming from the resolvable fission resonances of Pu-239 and Pu-241 at thermal neutron energies should be addressed. The aim of this study is to propose an effective solution to ameliorate

The fast periodic pulsed reactor utilizes rotating reflectors to introduce periodic varying reactivity, making the core generate continuous power pulses at a certain frequency. The pulse parameters under stable operation are extremely important in measuring the performance of the reactor. The critical search method of calculating the pulse parameters is established, and in order to further study the

It is a critical safety issue of thermal fatigue in the core outlet of sodium cooled fast reactors (SFRs) due to the mixing phenomenon of coolant with different temperatures flowing from different assemblies. In this paper, the large eddy simulation (LES) method was employed to study the coolant mixing phenomenon of a simplified assembly head model of SFR. The coolant velocity and temperature fields

A new multiphysics simulation suite has been created to model Liquid Metal-cooled Fast Reactors (LMFRs). LUPINE: “LMFR Utility for Physics Informed Nuclear Engineering” is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified PN (SPN) neutron transport equations and multiphysics models to simulate thermal hydraulics and thermal expansion. Results

During a decommissioning of nuclear facilities, rapid in-situ methods are highly desirable to support the characterization of material and sites. Mathematical efficiency calibration is suitable for in-situ measurements due to its ability to determine the efficiency for any geometry arrangement without the need to procure expensive calibration standards. During the characterization, the geometry of

As part of the Versatile Test Reactor (VTR) program, extensive validation work is being performed to demonstrate the fidelity of the Argonne Reactor Code (ARC) software package presently used for the core design studies. It is anticipated that this software will be also used for initial operations although additional software may be introduced at a later time to supplement or replace some components

The bonding mode of NITE-SiC matrix and TRISO particles has a great impact on the fracture mechanisms of the Fully Ceramic Microencapsulated (FCM) fuel. Combination strength and interface debonding characteristics are of great significance when evaluating the thermo-mechanical behavior of a FCM fuel. In this study, a finite element method (FEM) model for TRISO-matrix debonding characteristics analysis

Heat transfer in the rod bundle is augmented by the mixing vanes on the spacer grid. We conducted a computational fluid dynamics (CFD) simulation with three isotropic turbulence models – standard k–ε, realizable k–ε, and SST k–ω models – to investigate the relationship between heat transfer and turbulence behaviors downstream a simulant spacer (with four vanes) in a single tube under single-phase flow

Recently, a new computer code, ISOBURN (ISOtopes BURNup solver), has been developed supporting both standalone and coupled-with-MCNP modes. The focus of this paper is on the features of ISOBURN in the standalone mode. At first, we propose a flux normalization method based on a semi-empirical analysis of power-generation components to find the constant flux that could deliver demanded average power

A negative fuel temperature reactivity coefficient is an important reactor physical parameter that indicates the inherent reactor operating safety. However, existing Monte Carlo perturbation methods still cannot accurately and efficiently predict the derivative of the k-eigenvalue with respect to temperature. The main difficulty lies in calculating the derivatives of the microscopic cross-section with

This paper provides an overview of the capabilities of a new toolset that is being developed at Canadian Nuclear Laboratories for modelling advanced non-water cooled small modular reactor concepts. Throughout the study, a technology agnostic approach was adopted. The integrated toolset includes computational fluid dynamics, neutronics and system thermalhydraulics codes to handle a range of multi-physics

When adjusting nuclear data with integral experiments, care must be taken that spurious adjustments are not made by assimilating poorly characterized integral parameters. If there are unaccounted for biases or poorly estimated uncertainties in the calculated and experimental values for an integral parameter, the Bayesian data assimilation may adjust the nuclear data in a manner that does not reflect

A variety of modern Cyber-Physical Systems (CPSs) are distributed and asynchronous systems that are becoming the backbone for smart infrastructures and systems such as smart grids, power plants, medical appliances, social robots etc. The weaving of cyber components with the physical system improves resource utilization and system reliability. However, it makes the CPSs significantly vulnerable to several

This paper presents an integrated study about some accident-tolerant fuels compared to the standard fuel in trying to find an alternative fuel that enhances the safety, competitiveness, and economics of commercial nuclear power. The neutronic, thermal–hydraulic and solid mechanics analysis of Uranium Nitride (UN) and Uranium Carbide (UC) were performed in a single fuel assembly of the VVER-1000 nuclear

In this paper, a new adaptive genetic algorithm (AGA) is proposed for the loading pattern optimization (LPO) of the pressurized water reactor (PWR) core. The tournament selection, the two-point crossover and the mutation based on randomly swapping positions between two fuel assemblies (FAs) are applied. New calculation formulae are introduced to adjust effectively the crossover and mutation probabilities

The harmonization within the requirements concerning release from nuclear facilities has been a main issue in recent years. Implementation of the requirement of avoiding accidents that could result in large or early radioactive release to the environment, introduced also in the Council Directive 2014/87/Euratom is in progress in EU member countries, but the directive does not give guidance on practical

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR

A meta-stable effect on a subcooled critical flow in a nozzle and an orifice was investigated. For an investigation of the meta-stable effect, two homogeneous nucleation models, e.g., the Alamgir and Lienhard’s model and the Elias and Chambre’s model, and one conventional isentropic model, e.g., the Henry and Fauske’s model, were compared with respect to a previous test data performed by Sozzi and

The purpose of the present study is to develop the neutron diffusion solver and neutron noise simulator based on the Boundary Element Method (BEM). The 2-D, 2-G neutron/adjoint diffusion equation and corresponding neutron/adjoint noise equation were solved using the mentioned method. The developed neutron static and noise simulator based on the finite element method gives accurate results when the

With the development of digital information technology, nuclear energy systems are developing in the direction of intelligence and unmanned, which requires a higher demand for its safety, such as autonomous fault diagnosis. At present, the network structure model used in fault diagnosis usually needs professional design, which is time-consuming and labor-intensive, and the efficiency is low. To solve

One dimensional space nuclear reactor system analysis code—SIMCODE is developed and five typical transient operation conditions are calculated, including the system startup, the loss of flow accident, the loss of heat sink accident, the reactivity insertion accident, and the small loss of coolant accident. These conditions are part of the key performance of the space nuclear reactor system and are

The JANUS Phase 1, Phase 2 and Phase 4 shielding experiments performed in the 90’s in the NESTOR reactor aimed to study deep neutron propagation in mild steel, stainless steel, sodium and boron carbide. Together with the other JANUS phases (up to Phase 9) they represent invaluable experimental database for improving nuclear data for fast reactor applications. Several detectors were used to measure

In this research, the numerical simulation and analysis of flow behavior in 5 × 5 fuel rod bundle with consideration of spacer grid has been conducted. ANSYS FLUENT 19.0 with the consideration of the transition SST (Shear-Stress Transport) with SAS (Scale-Adaptive Simulation) turbulence model is used for the numerical simulation. The obtained simulation results are reasonable compared with the experimental

As a key device influencing the safety and the economy of nuclear power plants simultaneously, the main feed water regulating valve (MFWRV) controls the flow rate flowing to the steam generator. In this work, modal and structural analysis is conducted under different loading conditions to investigate vibration characteristics and mechanical characteristics of MFWRV. The results show that the combination

In designing a thermal–hydraulic facility simulating a floating nuclear power plant, the scaling of dynamic motion is necessary. As natural circulation under motion is a significant phenomenon in floating nuclear reactors, a scaling analysis for single-phase natural circulation was performed, and similarity criteria were derived for inclination, frame acceleration, and rotational motion conditions

Very high temperature reactor (VHTR) is a Generation-IV (Gen-IV) type of reactor which could nominally operate at a high temperature. In traditional two-step core neutronics calculation method of VHTR, the high precision multigroup cross sections are required. Therefore, a new Monte Carlo based generation scheme for multigroup cross sections of pebble type VHTR, based on open source code OpenMC, is

The paper presents a novel approach to sensor position optimisation based on compressed sensing (CS) and its application to neutron flux estimation in a nuclear reactor. An online flux mapping system (FMS) with a suitable algorithm is needed to continuously monitor neutron flux distribution from the measurements of in-core detectors in large nuclear reactors. In the present work, sensor positions are

The purpose of this work was to guide the design of the suspended control rod in a sodium-cooled fast reactor. Retaining the dimensions and main flow characteristics of the prototype device, a hydraulic model of the passive shutdown device was designed. An experiment was carried out on the critical flow of the upper working position and the lower position (accident condition). The critical flow rate

The Helical coiled tube Once-Through Steam Generator (H-OTSG) is key equipment in the High-Temperature Gas-cooled Reactor (HTGR) system, where the fission energy is transferred from the helium side to the water side, together with the complicated heat transfer and phase change phenomenon. In detail, the water is heated to the water-steam two-phase flow by the hot helium and further to the superheated

The creation of an international shielding benchmark database was presented in 1988 at the International Reactor Shielding Conference (ICRS7) in Bournemouth, UK. M. Salvatores was among the authors of the proposal and had promoted and contributed to the project since the first initiatives, showed continued interest and encouraged the development of the database. He was Chairman of the Committee of

All over the world, the nuclear reactor technology has been in the critical stage of transformation and upgrading. Also, there have been the higher requirements for the key technical indicators, such as the power level, the miniaturization and the security. According to analyzing the engineering requirements of typical nuclear reactor devices, the authors demonstrated the scientific problems and the

The Monte Carlo photon transport code IMPC-Photon is developed for serving the software development of designing advanced nuclear energy system such as Accelerator Driven Advanced Nuclear Energy System (ADANES) in Chinese Academy of Sciences. For photon physics, four main photon-atomic interactions and three secondary processes including atomic relaxation, electron-positron annihilation, and thick-target

In this study, a three-dimensional (3D) transient computational fluid dynamics (CFD) model is presented to investigated liquid entrainment in gas–liquid flow through vertical T-junction of fourth stage automatic depressurization system (ADS-4) in AP1000. A specialized CFD model for the liquid entrainment phenomenon study, which is based on coupled Eulerian-Eulerian VOF (volume of fluid) formulation

A Seismic Probabilistic Safety Assessment (SPSA), unlike other internal or external event analyses, has a very high probability of basic events when a strong earthquake occurs. Because of this high probability, the Core Damage Frequency (CDF) might be overestimated by the Rare Events Approximation (REA) and Minimal Cutset Upper Bound (MCUB) method for general quantitative analysis. To compensate for

The improved subgroup method adopts a fine energy mesh to handle the resonance interference effect, then the neutron slowing-down equation is solved for group condensation to obtain the multigroup cross section. A large calculating burden is raised in this process. In this paper, an acceleration scheme by the multigroup kernel method is investigated on both GPU and CPU platform. First, the subgroup

The decommissioning is the last stage in the shelf-life period of any nuclear or radioactive facility. The goal is the controlled removal of the source term in a way that the future use of the buildings decommissioned does not entail any radiological risk for both people and environmental according to a radiological criteria regulator approved. MARSSIM is the methodology followed by many countries

To investigate the effects of combining two survey units quantitatively in the final status survey of MARSSIM for a decommissioning facility, an expression for the sample size of combined survey unit was developed, which consists exclusively of related parameters of the two survey units. The simulation results using the expression showed that all combining cases considered always result in a decrease

The presented paper discuses a study of core degradation behavior where a Medium Break Loss of Coolant Accident (MB LOCA) with a150 mm internal diameter along with station blackout (SBO) was the initiating event. One of the goals of this work was the examination of core degradation at different primary side break locations. Two simulations of MB-LOCA scenarios were performed. The first analysis was

The balance between sustainability, energy security and affordability are important trade-offs to consider in decarbonizing the current energy system. The two practical alternatives for moving forward seem to be either reducing the energy storage costs to enable deployment the intermittent renewables on a large scale or developing an affordable and more flexible nuclear power. A small modular boiling

A numerical simulation is used to analyze the pellet-cladding mechanical interaction (PCMI) of a heavy water research reactor (RR) fuel rods under power ramps. A mechanical simulation that best reproduces the thermomechanical behavior of the pellet-cladding system is a major industrial challenge for engineers because it involves defining operating rules that are not too restrictive but that guarantee

The elevated temperatures and corrosive environment typical of molten salt reactors’ operating conditions may pose challenges to flow and temperature measurements. Experimental measurements of these thermal–hydraulic parameters are paramount to advance the knowledge and understanding of the molten salts’ behavior, and are necessary for the validation of system-level and computational fluid dynamics

A nuclear reactor is a complex, nonlinear, and time-varying system. External disturbances and uncertainty due to neutronic and thermal-hydraulics parameters variation contribute to reactor control challenges and safe operations. Thus, this paper presents a hybrid control system for a nuclear reactor core power control in the presence of a matched disturbance and uncertainties. The hybrid controller

A 3D MCCI code, CORCAAB, has been developed to account for uncertainty factors in the long-term molten corium–concrete interaction including changes in the three-dimensional cavity contour, which affect the convective heat transfer of the molten debris and late ablation behaviors. In this work, a 3D concrete ablation front tracking algorithm was formulated using the piecewise linear interface calculation

The semi-analytical model (a.k.a. the tube-in-channel model) needs less empirical inputs and considers interactions between tube movements and cross-flow variables, and is well used in fluidelastic instability (FEI) threshold prediction of tube bundles. However, despite numerous flexible tubes in engineering practice, tube-in-channel models with more than 7 tubes are not yet studied. In this paper

In this paper we report validation studies of the Monte Carlo code OpenMC by the ASTRA critical facility benchmark. A high-fidelity reactor model is built using the regular lattice coated fuel particle and the body centered cube pebble spheres representations. Two nuclear data libraries, ENDFB-7.0 and ENDFB-7.1 are considered to investigate the validity of nuclear cross section libraries. In addition

The applicability of the photofission reaction ratio (PFRR) method to identify high-enriched uranium is studied by switching the photon source from the Gaussian spectrum to a bremsstrahlung spectrum. The combination of 6 and 11 MeV Gaussian photon energy was used in previous study for applying PFRR method because the photofission cross sections between 235U and 238U at these two energies differ significantly

Japan Atomic Energy Agency has developed a new nuclear data processing code, named FRENDY, to generate ACE-formatted files from various evaluated nuclear libraries. A code-to-experiment verification of FRENDY processing was carried out in this study with criticality benchmark assessments of the high temperature engineering test reactor. The ACE files of the JENDL-4.0 and ENDF/B-VII.1 was generated

This paper deals with the analysis of the grid generated turbulence on flow induced vibration response of the parallel triangular tube bundle with pitch ratio of 1.54. Turbulence characteristics of the upstream flow can highly influence the stability behavior of a single flexible as well as multiple flexible tubes in early rows of tube bundle. Experimentation have been performed in the wind tunnel

Critical flow has significant impact on the reactor safety. Critical flow models are considered as an important module in all system thermal–hydraulic (STH) codes, with several optional models. But each model has its own scope of application. Consequently, the STH simulation results largely dependent on the model choice, i.e. on the user effect. A full-range critical flow model may be used to deal

Direct numerical simulation (DNS) is performed to study turbulent heat transfer in Poiseuille-Rayleigh-Bénard (PRB) flows with low Prandtl numbers in this article. The mesh is Cartesian and a highly accurate finite difference sixth-order compact scheme is chosen to discretize the incompressible Navier–Stokes equations to perform DNS. Liquids with a fixed Richardson number of 0.25 and four different

An experimental study of natural circulation flow instability has been carried out to investigate instability modes at low pressure. Water in the primary loop is heated in a single macro-channel and condensed in a tube-shell shape condenser by circulating coolant in the secondary loop. Two kinds of flow instability modes are identified: flow instability induced by flow boiling in the heated channel

Nuclear regulatory guidelines require all nuclear facilities to establish sufficient prevention and response capabilities against cyber-attacks. In this study, an integrated cyber-attack response process at NPPs is analyzed. The integrated response process includes a safety-response security analysis that enacts a safety response against unperceived damage, as well as against potential damage. To help

Common family of numerical methods such as finite element and finite difference involve the discretization of domain into regular grids or meshes. Mesh generation and re-meshing process, however, for obtaining required accuracy are complicated and time-consuming. Recently, Mesh Free (MFree) methods have been developed to overcome the drawbacks of such conventional numerical methods. In the present

Several reasons such as no free lunch theorem indicates that any learning algorithm in combination with a specific feature selection (FS) technique may give more accurate estimation than other learning algorithms. Therefore, there is not a universal approach that outperforms other algorithms. Moreover, due to the large number of FS techniques, some recommended solutions such as using synthetic dataset