When damaged core materials are relocated to the reactor lower head with penetrations during a severe accident, the loss integrity of the reactor pressure vessel caused by ejection of a penetration tube will occur earlier than the global failure of the reactor lower head. The overall aim of this study is to review and improve the previous contact pressure evaluation models. In this part of the study

The prevailing state of knowledge of two-phase flow in complex systems, as in the nuclear field, usually leads a certain degree of ad hoc calibration of computational models and the consequent inevitable subjectivity in analyses and assessments. The UPC-ANT Uncertainty Analysis methodology for code assessment attempts to maintain scientifically desired features of models and analyses, by crediting

The US NRC system code TRACE has been modified at PSI for application to liquid- metal-cooled reactor. An unprotected loss-of-flow-without-scram test performed at the Fast Flux Test Facility (FFTF) provides an opportunity to enhance the validation base of TRACE to transient analysis for sodium-cooled fast reactor (SFR). The FFTF primary system model was created with TRACE and initial core flow distribution

DARWIN/PEPIN2 is a French inventory code that has been developped at CEA since the mid-1990s for fuel fission reactor, material neutron activation and material radioactive decay calculations. It takes into account both R&D and industrial application requirements. Today, DARWIN/PEPIN2.4.x is the last major release and deals with nuclear data uncertainties propagation to some quantities of interest outputs

The effect of feedback reactivity on the safety of a fast reactor has been studied based on the asymptotic approximation for unprotected transients. The asymptotic approximation adopts the steady state heat transfer equation for calculating temperatures of coolant and fuel with the use of the relevant core power and mass flow. The feedback reactivity can be approximately estimated from the temperatures

The method of discrete ordinates is investigated for neutron noise simulations in the frequency domain. For this purpose, the solver NOISE-SN is developed and used to simulate two neutron noise problems that are respectively derived from the two-dimensional systems described in the neutron transport simulation benchmarks C4V and C5G7. In the first problem based on the C4V system, NOISE-SN is compared

Crucial to the development of molten salt reactor (MSR) designs is the application of multiphysics codes to model the tightly coupled neutronics and thermal-hydraulics behaviour of the liquid fuel. However, the verification and validation of such codes is not a trivial task, in particular for fast reactor designs, where no experimental data are available. In absence of experimental data, a benchmark

The present work examined the feasibility of utilizing Mixed Uranium–Plutonium Oxide (MOX) fuel in the VVER-1000 reactor core. In this study, about 30% of Uranium Oxide (UOX) fuel assemblies of the original core are replaced by MOX fuel assemblies. The effect of MOX fuel on the safety parameters is studied by using the MCNP6 code and the nuclear data library (ENDF/B)-VII.1. The safety parameters examined

The SWIFT-RIMPUFF is composed of the Stationary Wind Fit and Turbulence model and the Risø Mesoscale PUFF model, which has explicit building modules for atmospheric dispersion modeling. A comprehensive evaluation of SWIFT-RIMPUFF was performed to the vertical profiles of wind speed, horizontal wind and concentration fields based on six wind tunnel experiments of the Sanmen nuclear power plant site

The main advantage of using Nodal Integral Methods (NIM) in solving partial differential equations (PDEs) is that they lead to an accurate solution over relatively coarse meshes. Due to the use of the transverse integration procedure in the derivation, most of the NIMs developed were restricted to PDEs with isotropic diffusion terms only. Recently, the NIM has been extended to arbitrary geometries

The covariance between sensitive cross-sections is adjusted in the cross-section adjustment method even if these cross-sections are originally independent of each other. On the other hand, the experimental error and calculation error are also sensitive to the results, thus the covariance among not only cross-sections but also errors of experiment and calculation can be adjusted. However, the derivation

Removing the solidified core melt from damaged reactors by drilling or cutting requires knowledge of both mechanical properties and its microstructure. The ingot of the light-water power reactor prototype corium with a different blend composition containing ZrO2, UO2 and Zr was obtained in this work. The microstructure has been studied and the mechanical properties of the light-water power reactor

Nuclear data are the main uncertainty sources for modern neutron transport calculations. However, complete correlation matrices among different nuclear data are not presented in the current evaluations. In order to completely consider that correlation between nuclear data, the present work propagates the covariance matrix of continuum n+56Fe reaction model parameters to the effective neutron multiplication

The generation of real nuclear pulse signals need a detector and a radiation source. In order to replace some expensive detectors or harmful radiation sources, and precisely evaluate a nuclear detection system, a monte Carlo virtual detector (MCVD) which could simulate different types of detectors and radiation sources was designed to generate nuclear pulse signals. The generated nuclear pulse signals

The new ENDF/B-VIII.0 evaluated nuclear reaction data library is improved relative to the previous ENDF/B-VII.1 library because of using new evaluated data for neutron and photon reactions on certain nuclides. To get a more comprehensive impression for shielding calculations on the latest neutron and photon data library, and to provide reference for the selection of the evaluated nuclear data libraries

A new approach is proposed to evaluate the safety of molten-salt reactors (MSRs), using advanced modeling and simulation (M&S) tools. This approach augments the Monitoring and Inspection (M&I) concept for fuel qualification by enabling computation of the change in critical safety parameters as a result of altered thermo-physical fuel properties. This work uses newly developed capabilities in the MOOSE

In this paper, we derive a mathematical formulation of the higher order adjoint NEM-M2B2 equations by preconditioning the nodal interface neutron currents equations of the forward equations system, and by using the Lagrangian Multipliers analysis method. In the NEM-M2B2 system of equations, the quadratic transverse leakage approximation is used to model the leakage of neutrons between each node in

In this paper, fracture properties of Cr-coated Zircaloy cladding, which has initial oblique crack in the coating, have been investigated. The von Mises stress distribution caused by different modes of crack has been calculated. By comparing the stress intensity factor with the fracture toughness of material, the fracture mechanical behavior of Cr coating has been obtained. When the reactor is in normal

The new paradigm of nuclear reaction evaluation postulates that differential and integral experiments along with the reaction models should be used in concert to produce the evaluated data files. The results of a previous assimilation project are summarized as a proof of principle that adjusting reaction model parameters to the results of integral experiments is feasible. The new paradigm requires

Nuclear reactors are built with many inherent and engineered safety features with defence in depth philosophy, to maintain integrity of core during normal as well as accidental conditions. However, a severe accident scenario such as hypothetical core meltdown has to be considered in safety analysis of the plant, towards demonstrating a safe and coolable geometry of degraded core. Progression of corium

The reactor cavity cooling system (RCCS) is a common reactor safety system in high-temperature gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel (RPV) by radiation (∼80%) and natural convection (∼20%). For the simulation of accident scenarios of HTGRs, intermediate fidelity and system codes models must be employed to limit the models’ execution time. While an accurate quantification

A multiparameter representativity approach for the design of Advanced full MOX BWR core physics experimental programs in ZPR was developed. The approach is based on sensitivity analysis of integral parameters to nuclear data, and correlations among different integral parameters, largely used for Fast Reactor Applications, in particular by Salvatores et al at the French Atomic and Alternative Energies

As dynamic loading caused by turbulent flow may affect the integrity of reactor vessel internals (RVIs), characterizing relevant features and establishing an appropriate assessment procedure are vital. This study addresses influence of turbulent flow as a problem obstructing the safe operation of RVIs. Numerical models are developed and optimized by comparing preliminary analysis data with theoretical

Flow instability can cause mechanical oscillations in the equipment as well as periodic changes in the thermal stress, both of which have a great impact on the safety of the reactor. In addition, flow instability can interfere with the control system, making it difficult to capture stable parameters. In this study, to examine the change in the stability of a parallel channel-loop system with natural

High temperature heat pipe (HTHP) employs alkali metal or its alloy as working fluid. Due to the large latent heat of its working fluid during phase change heat transfer, HTHP has the advantages of high temperature uniformity, high working temperature and high intensity heat transfer capacity, which has broad application prospects in the field of HTHP cooled nuclear reactors and the thermal protection

Molybdenum is considered as one of the substitutes for zirconium in making fuel cladding for safe reactor designs. The isotopically modified molybdenum (IMM) should be used instead of natural molybdenum to achieve higher energy efficiency, reduce uranium consumption, and mitigate the impact on the nuclear fuel cycle. This task can be fulfilled by using the same technology of gas centrifugation for

For the design verification and licensing of newly pressurized water reactors (PWRs), many integral test facilities were conducted. From the above integral test facilities, the obtained experimental results were adopted in the assessment and development of safety analysis codes. In this study, scaling analysis and design parameters of integral test facilities were summarized from the perspective of

TRISO coated particle fuel exhibits statistical variations and large uncertainties in physical configurations and material properties from particle to particle. Besides, the dependence of mechanical behavior and failure mechanism of the coating layer on multiple variables (such as density, porosity, anisotropy, fluence and temperature) further increase the complexity of TRISO fuel performance analysis

The Gas-cooled Fast Reactor is an advanced concept selected to be part of the fourth generation of nuclear reactors. Before its development, the construction of the experimental demonstration reactor ALLEGRO is planned. Given the current interest in this innovative system, a neutronic study of the ALLEGRO reactor core is performed using the Monte Carlo Serpent calculation code and the deterministic

The errors in pin-homogenized multigroup calculation (PHMGC) from various sources are thoroughly analyzed to find the optimal neutronics solution method. The dependence of the errors due to the use of pin-homogenized multigroup cross sections on the energy groups are investigated first to clarify the need for proper equivalence factor and more groups in pinwise calculation. Various transport approximations

Part I of this study introduced a novel methodology for the optimization of energy group structures to be used in diffusion calculations of Sodium cooled Fast Reactors. The goal of the optimization is to speed up calculations, particularly in transient analyses, while maintaining an acceptable accuracy of the results. The proposed methodology is based by either direct or simulated annealing search

In this study, a modified two-dimensional geometry model was employed comprised of metal cladding, a metal matrix, and dispersed fuel particles. And an analysis method was established to simulate the thermal–mechanical behavior of dispersed plate-type fuel under the Reactivity Insertion Accident (RIA). Key fuel properties including thermal performance, mechanical behaviors and irradiation effects were

In this study, preliminary neutronic calculations were performed in order to provide a basic design of a prismatic-core advanced high temperature reactor (PAHTR) that adapts a mixture of lithium fluoride and beryllium fluoride (Li2BeF4, known as FLiBe) as a coolant. This work investigates the effectiveness of utilizing an integral fuel burnable absorber (IFBA) as an additional coating layer for the

Nuclear fuel performance analysis is fundamental for the design of pressurized water reactors (PWRs) with great safety concerns. The analysis of a fuel rod involves many complex processes, including heat transfer, mechanics, material, mass transfer and chemistry interaction. This paper reviews the basic frame and recent progress of fuel performance analysis of PWRs during normal operation and accidents

In recent decades, separation of stable isotopes due to their substantial role in human health has been widely increased. The present research deals with square cascades optimization in order to separate the 123Te by the Gray Wolf Optimization algorithm (GWO). The separation of 123Te has significant application in medical science, and production of radioisotopes. In this study, attempts have been made

Due to the excellent properties of high temperature strength, creep resistance, lower thermal expansion and better irradiation resistance, the silicon carbide is considered as a substitute for the Zircalloy cladding. In this paper, the thermal and mechanical properties of SiC cladding are reviewed firstly. Then the thermophysical properties and mechanical behavior model of the thorium-based fuel and

The study of particle discharge flow is significant in industrial processes,especially wet cohesive particles. In this paper, the discrete element method (DEM), cohesion model and the viscous force model are used to study the influencing parameters of the discharge flow characteristics of cohesive system in three-dimensional packed bed. The cohesion between particles is described by liquid bridge model

This work is aimed at developing an advanced model of the Slovak VVER-440/V-230, reactor in the MCNP code, to calculate induced activity in ex-vessel structural components and to validate obtained results to achieve an effective calculation tool for decommissioning purposes. A survey of the V1 NPP operating history revealed that after 12 fuel cycles the shielding assemblies were placed at the core

Safe disposal of nuclear waste is key to sustainable use of nuclear energy. Exploring safe and reliable disposal methods for nuclear waste has become a research hotspot. The phase distribution of uranium in matrices obtained by co-melting basalt and uranium-containing Al2O3 was examined for the development of a potential material for radioactive waste immobilization. The co-melting of basalt with UO3-containing

The radionuclide composition of, and emitted radiation in, spent nuclear fuel from the future MYRRHA facility have been studied using depletion simulations to understand potential consequences for safeguards verification using non-destructive assay. The simulations show that both the gamma-ray and neutron emission rates in spent MYRRHA assemblies are lower than in spent PWR UO2 and MOX assemblies.

Mitigating of molten corium in severe accidents is a vital concern in safety of Nuclear Power. Plants (NPPs). As one of the main features, core catcher is provisioned in the design of NPPs. In order to appropriately design an ex-vessel core catcher, numerous core damage (CD) sequences with vessel breach are required to be studied and analyzed, which requires massive efforts. Therefore, this study intends

Submerged gas jet plays a very important role in many industries occasion. It often causes load of pool structure induced by flow oscillation. Exploring the mechanism and key influencing factors of submerged non-condensable gas jet oscillation characteristics is of great significance to ensure the safe operation of the equipment. In the present work, the phenomenon of flow oscillation for non-condensable

A new computational method, by incorporating wavelets in the improved quasi static (IQS) scheme, is developed and presented here for the analysis of CANDU PHWR and TWIGL reactor transients. According to this method the time dependent neutron diffusion equation is split into amplitude function and shape function. The amplitude function is solved using the Haar wavelet and its solution is used in the

This article presents the current state of knowledge on beryllium, its nuclear properties and reactions, based on experience gathered in multiple material testing reactors or beryllium manufacturers. The main focus is given to reactions of high importance due to their impact on the reactor's operation and safety, in a particular buildup of both tritium and neutron absorbers - so-called poisons. Few

In this paper, a new pattern bundle with a smaller wire diameter of peripheral rods was proposed. By decreasing the wire diameter of peripheral rods, the flow area of peripheral sub-channels is reduced diverting more coolant towards the central sub-channels. This would lead to a more uniform sub-channel temperatures distribution. To analyze the flow and heat transfer performance of the new bundle,

The temperature distribution is an important issue for the design and safety assessment of the repository for disposing high-level radioactive nuclear waste (HLRW). In order to analyze the heat conduction property near the HLRW canister accurately, the solution of temperature distribution is divided into two parts. Two mathematical models were established respectively, and the corresponding analytical

As an effective data pattern extraction method, machine learning is widely used in the field of nuclear power industry control, and has a great application. In order to solve the problem that there is a serious lack of operation data set for analyzing various variable conditions or accident conditions, this paper uses the maximum likelihood estimation and the least square estimation to prove that,

Uncertainties estimation is essential for a sound assessment and a clear definition of safety margins, which in turn are crucial for licensing approval of new nuclear reactors by safety authorities. Moreover, uncertainties are also matter for target accuracy requirements (TARs), whose achievement could disclose new or enhanced opportunities. The concept of target accuracy requirements has been over

The implementation and optimization strategies of the neutron tracking algorithms in the GPU-based continuous-energy Monte Carlo (MC) code PRAGMA are presented. The strategies consist of 1) a mixed precision technique, 2) an event-based tracking algorithm, 3) the history-based tracking algorithm improved for GPUs by imposing transition limits, 4) the region partitioning and energy sort schemes to vectorize

A series of experimental study on natural circulation pressure drop oscillations (PDO) under rolling motion are carried out. PDO under rolling motion is a chaotic motion, the thermal–hydraulic mechanism and evolution route are analyzed by maximal Lyapunov exponent and spectrum analysis. When rolling motion has different periods with PDO, composite oscillations are formed. The evolution route can be

This paper presents a novel approach for the modelling of the reactor kinetics of subcritical systems. The new model, named multi-generation point kinetics, shares some of the features of the classical point kinetics, such as the factorization of the solution in shape / amplitude and the projection of the reactor kinetic equations to derive a lumped parameter model. Unlike classical point kinetics

An iterative regularization method is proposed for neutron spectrum unfolding of the liquid organic scintillator EJ-301. The Arnoldi iteration is used to generate Krylov subspaces. GMRES (Generalized Minimal Residual) combined with Tikhonov regularization is used for neutron spectrum updating. The restart method is applied to speed up the convergence as well as non-negativity constraints. The Monte

A mathematical and computational framework is introduced to assess and improve the fitness of multivariate distribution models used in combination with the Bayesian data assimilation framework MOCABA. This is achieved by expanding the distribution model using invertible vairable transformations. Such model expansions enable us to generalize the basic MOCABA framework in order to make it applicable

As the need for simulations of multi-dimensional behavior in the safety analysis is increasing, efforts have been devoted to developing a three-dimensional module for the conventional system analysis codes. After having developed a one-dimensional system analysis code, MARS, the Korea Atomic Energy Research Institute (KAERI) began to develop the CUPID code to address the need for a multi-dimensional

In this study, compactness improvement capacity of a self-powered efficiency booster is analyzed for natural circulation systems used in integral small modular reactors (SMR). Based on thermoelectric principle, the booster converts heat energy into electricity directly, which is used to produce additional driving force for natural circulation flow and heat transfer. Thus, flow and heat transfer are

The cumulative and independent yields of various fission products within the mass range of 83–155 have been measured in the thermal neutron induced fission of 238Pu by using an off-line γ-ray spectrometric technique. From the cumulative yields, post-neutron mass chain yields were obtained by using the charge distribution correction. From the mass yields data, the full width at tenth maximum (FWTM)

The present work exposes a communicating vessels closed system pre-conceptual design for a small Molten Salt Reactor (MSR). The fuel volume is kept in a critical configuration by using forced convection of an immiscible non-reactive neutron absorber with a higher density than that of the fuel. The absorber is pushed to make the fuel flow into the core. The fuel interfaces with the absorber that fills

The condensation heat transfer characteristics of steam with the non-condensable gases (NCGs) in the tubes bundle channel are quite different from those in the simple geometry structure, such as the single tube and flat plate. In the present work, to evaluate the thermal characteristics of steam condensation with NCGs in the vertical tubes bundle channel, the numerical model of steam condensation was

In pebble-bed high temperature gas-cooled reactors (HTGRs), the outside graphite reflector generally contains void regions, including the boring channels for control rods, absorber balls and coolant flow. Deterministic reactor physics calculation of HTGR is usually based on regular geometries, which requires pre-homogenization to the materials with void regions. However, the volume-weighted homogenization

Reactor Cavity Cooling Systems (RCCSs) with passive safety features use the atmosphere as a coolant, which cannot be lost; however, the drawback is that they are easily affected by atmospheric disturbances. To realize the commercial application of two types of passive RCCSs, their safety must be evaluated, i.e., they must be able to remove heat from the reactor at all times and under any conditions