We present the new iterative method lpCMFD-SOR, which combines the linear prolongation coarse-mesh finite difference (lpCMFD) scheme with the method of successive overrelaxation (SOR) for neutron transport source iteration (SI). The lpCMFD method is the latest coarse-mesh finite difference (CMFD)–type acceleration scheme and is unconditionally stable and more effective than the standard CMFD method

The safe management of radiation sources and wastes is one of the most important elements in operating nuclear power plants (NPPs). Safe management requires periodically measuring radiation during the operation and decommissioning of NPPs, but it is impossible for radiation management systems to cover all areas, and it may be necessary for a person to measure radiation directly where the risk is high

The CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) reactor concept was proposed to overcome the disadvantages of current reactor technologies. In this study, a Monte Carlo–based procedure is developed for quantitative comparison of burnup performance and neutronic characteristics between lead bismuth eutectic (LBE)–cooled and sodium-cooled

The Scale Method was applied for analysis of experimental and theoretical prompt fission neutron spectra (PFNSs). This approach allowed us to demonstrate evidence from several experiments that had not been discussed before. The comparison of experimental and calculated data; the analysis of experimental PFNSs from neutron-induced fission reactions for 232Th, 233U, 235U, 238U, 237Np, and 239Pu; and

Recent development of the high-order, low-order (HOLO) method has shown promising results for solving thermal radiative transfer problems. The HOLO algorithm is a moment-based acceleration, similar to the well-known nonlinear diffusion acceleration and coarse-mesh finite difference methods. In this work, we introduce a new spatial-differencing scheme for the low-order (LO) system based on the corner-balance

At the Kyoto University Critical Assembly experiments on kinetics parameters are carried out at near-critical configurations, supercritical and subcritical states, in the thermal neutron spectrum made with a highly enriched uranium fuel. The main calculated kinetics parameters, the effective delayed neutron fraction (βeff) and the neutron generation time (Ʌ), are used effectively for the estimation

We report here the results of a measurement of the scattered versus unscattered neutron fluence on polyethylene determined via neutron activation of multiple natural indium foils from a deuterium-deuterium (D-D) neutron generator. The neutrons were produced by the High Flux Neutron Generator (HFNG) at the University of California, Berkeley, a specially designed source to maximize neutron flux on a

In this work, we attempt to overcome the “curse of dimensionality” inherent to neutron diffusion kinetics problems by employing a novel reduced-order modeling technique known as proper generalized decomposition (PGD). The novelty of this work is that it represents the first attempt at applying PGD reduced-order modeling to time-dependent multigroup neutron diffusion kinetics. The performance of PGD

This work is motivated by the need to solve realistic problems with complex energy, space, and angle dependence, which requires parallel multigroup transport sweeps combined with efficient acceleration of the thermal upscattering. We present various iterative schemes based on the two-grid (TG) diffusion synthetic acceleration (DSA) method. In its original form, the TG method is used with the Gauss-Seidel

Transient analysis for possible prompt supercritical accidents of fuel debris in the Fukushima Daiichi Nuclear Power Station is quite important. However, unlike solution fuel systems, there is little knowledge about supercritical transient analysis in fuel debris systems. In particular, reactivity feedback effects, which may have a significant impact on the results of the analysis, are important and

This paper presents the three new pin-resolved transient solvers of PROTEUS-MOC developed in a consistent way to the latest steady-state solver. A new transient fixed source problem (TFSP) solver was developed without relying on the isotropic approximation of the angular flux time derivative. A moving axial mesh scheme was also implemented to model the control rod movement accurately with coarse axial

Based on the angle and Euclidean distance similarity of the loading matrix, this paper reports on an improved principal component analysis (PCA) modeling method that is successfully applied to identify three different working conditions in the secondary loop of a nuclear power plant (NPP). First, a simulation platform of the secondary loop of the NPP is built in which three kinds of working conditions

The recently developed linear prolongation Coarse Mesh Finite Difference (lpCMFD) acceleration scheme, which employs a linear additive approach to update the scalar flux, has been shown to be more stable and effective than the conventional scaling-based Coarse Mesh Finite Difference (CMFD) method for accelerating the discrete ordinates (SN) neutron transport calculation using spatial finite difference

This paper discusses modeling of the performance of a VVER-type nuclear power plant using the Fractional Neutron Point Kinetics (FNPK) model. The modeling intent is to achieve a nonlinear system of fractional-order differential equations that are solved using SIMULINK by developing a scheme with the FOMCON Toolbox in the Matlab® environment. The model is shown to be identifiable, and the goodness of

Uncertainty analysis is a key element in nuclear power plant deterministic safety analysis using best-estimate thermal-hydraulic codes and best-estimate-plus-uncertainty methodologies. If forward uncertainty propagation methods have now become mature for industrial applications, the input uncertainty quantification (IUQ) on the physical models still requires further investigations. The Organisation

Integral systems test (IST) facilities, which are sometimes referred to as integral effects tests (IETs), play a key role in the design, assessment, and certification of innovative reactor designs. Data obtained using such facilities have been used to benchmark the best-estimate safety analysis computer codes used to evaluate nuclear plant safety. They have also been used to assess the effectiveness

Abstract We apply a nonlinearly preconditioned, quasi-Newton framework to accelerate the numerical solution of the thermal radiative transfer (TRT) equations. This framework was inspired by the unpublished method that has existed for years in Teton, Lawrence Livermore National Laboratory’s deterministic TRT code. In this paper, we cast this iteration scheme within a formal nonlinear preconditioning

The computational kernel in solving the SN transport equations is the parallel sweep, which corresponds to directly inverting a block lower triangular linear system that arises in discretizations of the linear transport equation. Existing parallel sweep algorithms are fairly efficient on structured grids, but still have polynomial scaling, P1/d + M, for d dimensions, P processors, and M angles. Moreover

Multigroup constants for deterministic methods that preserve the time-dependent physics of the neutron transport equations are derived. Alternative multigroup constant weighting spectra are discussed: (1) the fundamental k-eigenfunction, (2) the fundamental α-eigenfunction, and (3) a composite of several α-modes. To generate the fundamental α-eigenfunction for calculating the multigroup constants,

This technical paper presents an application of the GEneralized TwO Phase flow (GENTOP) model for phase transfer and discusses the submodels used. Boiling of a heated surface under atmospheric conditions is simulated by the multifield computational fluid dynamics (CFD) approach. Subcooled water in a generic pool is heated up first in the near-wall region leading to the generation of small bubbles.

A recent hybrid stochastic-deterministic calculation scheme using Monte Carlo–tallied group cross sections in a deterministic solver uses the best of both worlds for accurate and fast reactor agnostic transport simulations. However, neglecting the angular dependence of group cross sections induces large self-shielding errors in resonance groups, causing a large reactivity bias up to 300 pcm in light

Results reported in the literature have shown that the turbulence models currently implemented in computational fluid dynamics (CFD) commercial codes (e.g., ANSYS-CFX, STAR-CCM+, and FLUENT) have a tendency to overestimate thermal stratification and underestimate turbulent mixing when buoyancy effects become dominant with respect to momentum effects. Also, standard large eddy simulation models cannot

The mechanical vibrations of core internals such as fuel assemblies (FAs) cause oscillations in the neutron flux that require in some circumstances nuclear power plants to operate at a reduced power level. This work simulates and analyzes the changes of the neutron flux throughout a nuclear core due to the oscillation of a single FA without considering thermal-hydraulic feedback. The amplitude of the

Various flow regimes exist in a boiling water reactor (BWR) as the steam quality increases in the uprising coolant flow, from bubbly flow, slug/churn flow, to annular flow. The annular flow is characterized by the presence of a fast-moving gas core and the surrounding liquid film flowing on the conduit wall. In addition, entrained droplets can be observed in the gas core with ingested bubbles in the

This paper presents the three new pin-resolved transient solvers of PROTEUS-MOC developed in a consistent way to the latest steady-state solver. A new transient fixed source problem (TFSP) solver was developed without relying on the isotropic approximation of the angular flux time derivative. A moving axial mesh scheme was also implemented to model the control rod movement accurately with coarse axial

Thermally induced density wave instability (DWI) (Type-II) is an important phenomenon for two-phase flow industrial systems. Developing numerical tools and methods for the prediction of the DWI boundary is of importance in the design and safety of nuclear reactors. With the advent of computational fluid dynamics (CFD) in nuclear safety analysis, it is important to first verify the CFD results against

The current standard for computational neutronic analysis of nuclear power plants (NPPs) is the so-called conventional approach, which relies on few-group, coarse-mesh diffusion calculations. The recent evolution of computing clusters and computational techniques gives the opportunity to use codes that perform first principles–based multiphysics simulations, allowing high resolution of the calculated

Abstract Recently, a Eulerian-based two-fluid computational fluid dynamics (CFD) framework with a wall heat flux partitioning approach has been intensively investigated for departure from nucleate boiling (DNB) simulation under the U.S. Department of Energy–funded Consortium for Advanced Simulation of Light Water Reactors (CASL) program. Understanding of the DNB characteristics over a range of pressurized

During a loss-of-coolant-accident (LOCA) transient in a pressurized water reactor (PWR), water from the primary circuit is lost at the break. PWR designs are equipped with safety systems (SS) such as safety injection or accumulators to inject water into the primary circuit and prevent the core from being degraded. Depending on the size, position, and orientation of the break, a part of the safety system

The purpose of this technical note is to evaluate the discharged fuel of breed-and-burn (B&B) reactors. The discharged burnup in a B&B core can be high, and there is a concern that as decay heat increases, handling after a shutdown might be difficult. Because discharged fuels contain a number of plutonium nuclides, the potential for proliferation is also a concern. Moreover, radiotoxicity levels are

The study of the nuclear source term requires the computation of aerosol dynamics. Solutions to the aerosol general dynamic equation (GDE) are difficult to obtain by analytical or numerical methods when more realistic problems are considered. The direct simulation Monte Carlo (DSMC) technique is capable of simulating aerosol evolution reducing simplifications in the implementation of the aerosol GDE

A new single scattering adjoint transport capability was implemented in Framework for REsearch in Nuclear ScIence and Engineering (FRENSIE). The Evaluated Electron Data Library (EEDL) was used to generate new tabulated adjoint data. All adjoint data were generated using refined EEDL data and a unit-base grid policy. Verification and validation tests were performed for the adjoint electron transport

The n-p scattering angular distribution was measured with 14.9 MeV incident neutrons produced at the neutron facility of Ohio University. The traditional time-of-flight technique with neutron-gamma discrimination was applied for the measurement of the number and energy of scattered neutrons. The scattering angle varied from 20 to 65 deg (laboratory system) in 5 deg incremental steps corresponding to

A high-fidelity safety analysis method for pressurized water reactors (PWRs) is presented using a multiscale and multiphysics coupled code. Computational resolution of the conventional safety analysis can be greatly improved using this method in which the whole reactor vessel is modeled at a subchannel scale with around 5 million calculation meshes. Three-dimensional thermal hydraulics inside the reactor

The static correlations from RELAP5 and TRACE as well as the interfacial area transport equation (IATE) are benchmarked for flashing flow with selected cases from a recently published experimental data set. The static correlations are observed to severely overpredict the interfacial area concentration by several orders of magnitude, especially under low-pressure conditions. Although the one-group decoupled

This paper examines the potential of embedded intelligence (EI) and the resultant implications of secure EI (SEI) and integrated state awareness (ISA) as disruptors in the advanced nuclear energy systems market. To this end, it reviews the state-of-the-art of EI in nuclear and other industries; discusses design trends, benefits, challenges, and opportunities; and addresses the rigorous development

The Egyptian Second Research Reactor (ETRR-2) is a pool-type reactor, 22 MW thermal, with 27 fuel elements loaded with 60Co production facility in the most relative highest flux position for the production of 200 Ci/g specific activity. The production of this specific activity needs a very long irradiation time and continuity of operation to produce useful quantities of 60Co over a reasonable period

Monte Carlo (MC) codes are widely used for the accurate modeling of nuclear reactors. However, efficient inclusion of thermal-hydraulic (TH) feedback within the MC calculation sequence is still an open problem. The issue is emphasized when coupled MC-TH calculations are needed to model the burnup evolution using multiple depletion steps. Among the techniques proposed to solve this problem is the utilization

Performance enhancement of the spectral analysis method (SAM) for evaluating the real variance of local tallies from the partial current–based coarse-mesh finite difference (p-CMFD) feedback is verified and explained. In the SAM, on successive Monte Carlo (MC) cycles, the real variance is obtained from the cyclewise samples instead of an explicit evaluation of covariance. However, if the cycle correlation

This paper presents the first application of model calibration to neutron multiplicity counting (NMC) experiments for cross-section optimization that is informed by adjoint-based sensitivity analysis (SA) and first-order uncertainty quantification (UQ). We summarize previous work on SA applied to NMC and describe notable modifications and additions. We give the procedure for first-order UQ and Bay

Usually, simulation tools are validated based on experimental data considering a best estimate simulation case; however, there is no quantification of this validation, which remains based on rough expert judgment. This paper presents advanced validation treatment of the simulation tool OCARINa devoted to unprotected transient overpower (UTOP) accidents on two CABRI tests considering this time a Best

Abstract Accident tolerant fuels (ATFs) are being tested by different nuclear vendors and research organizations, and their introduction into the U.S light water reactor fleet is planned for the second half of the 2020s. Under the framework of the U.S. Department of Energy Light Water Reactor Sustainability (LWRS) Program, as part of the LWRS Risk-Informed Systems Analysis Pathway, research is being

Solving the radiative transfer equation with the discrete ordinates (S  N) method leads to a nonphysical imprint of the chosen quadrature set on the solution. To mitigate these so-called ray effects, we propose a modification of the S  N method that we call artificial scattering S  N (as-S  N). The method adds an artificial forward-peaked scattering operator that generates angular diffusion to the

Abstract Liquid salts have become more attractive as coolants for low-carbon power generation due to needs for high-temperature heat and affordable energy storage. Of particular interest are halide salts utilized in fluoride-salt-cooled high-temperature reactors, molten salt reactors, and high-magnetic-field fusion machines, as well as in concentrated solar power systems. Because of their high-temperature

As part of its design certification effort, NuScale Power has completed a series of low mass flux [<1000 kg/(m2‧s)] critical heat flux (CHF) tests for a wide range of pressures at Stern Laboratories in Canada. Earlier studies have demonstrated that under annular flow conditions, disturbance waves with circulating vortices traverse the rod surface. The disturbance wave slides over and significantly

The prompt neutron decay constant α in a steady-state subcritical system can be directly measured using a reactor noise analysis method such as the Feynman- α method. To reduce the nuclear data–induced uncertainty of keff for a target system, this study investigates the applicability of data assimilation techniques, i.e., the bias factor method and the cross-section adjustment method, based on a subcritical

The impact of various parameters in the coarse mesh finite difference (CMFD) acceleration method on overall convergence behavior is investigated through numerical calculations using the method of characteristics (MOC). Four parameters appearing in the CMFD acceleration with MOC, i.e., scalar flux distribution in flat flux regions (FFRFlux), the scalar flux distribution in CMFD meshes (CMFDFlux), homogenized

The serpentTools Python package is presented as a useful and efficient alternative for processing Serpent results. One positive attribute of Serpent is that many output files are exported directly in a MATLAB format, allowing for results to be loaded with minimal to no effort. However, some files for larger analyses may require immense amounts of memory to load and store all the data, leading to long

Recent advancements in computing power allow utilization of state-of-the-art direct numerical simulations (DNSs), coupled with interface tracking techniques, to perform fully resolved simulations of complex two-phase flows, such as flow regime transitions. Studying the highly resolved temporal and spatial information produced from these virtual experiments can advance our understanding of the phenomenon

The task of this investigation is to design a controller that has a strong robustness in various operating conditions. A new structure of state feedback assisted classical control (SFACC) that uses a differential lag compensator in the inner classical control loop is proposed to improve the robustness of original SFACC. The linear quadratic Gaussian with loop transfer recovery (LQG/LTR) at the plant

The capability to validate integral transient response models is of critical importance for licensing new reactor designs. The Kairos Power testing program has developed a methodology to design scaled experiments that predict the thermal fluid behavior of the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor (KP-FHR). Such experiments will be used as part of the assessment base of evaluation

Monte Carlo (MC) transport codes offer high-fidelity modeling of particle transport physics, but their high computational cost makes them impractical for many applications. For some applications such as multiphysics and depletion that use finely discretized geometries, a large portion of this computational cost is attributable to ray tracing. Neighbor lists are a well-known method for accelerating

The work presents validation of the TRAC/RELAP Advanced Computational Engine (TRACE) code for natural circulation two-phase flow in a vertical annulus. Natural circulation experiments were recently conducted for a vertical internally heated annulus at the Multiphase Thermo-Fluid Dynamics Laboratory at the University of Illinois. The experimental matrix consists of 107 experiments with system pressure

In the Monte Carlo method for criticality calculation, the convergence-in-distribution check of the sample mean of tallies can be approached in terms of the influence range of autocorrelation compared to the total number of generations iterated. In this context, it is necessary to evaluate the attenuation of autocorrelation coefficients (ACCs) over lags. However, in just one replica of calculation

This study introduces a radiation avoidance algorithm to help radiological occupational personnel (ROP) avoid high radiation exposure in a radioactive environment. The premise of this study is that ROP can be designated as a movable point in a two-dimensional radioactive scene with known radioactive sources. A trajectory of ROP is generated by the radiation avoidance algorithm based on an artificial

Alloying additions are introduced into U-Zr fuel in order to bind lanthanides (e.g., cerium) and prevent their migration to the fuel-cladding interface. Antimony (Sb) is being investigated as a candidate additive. The present study focuses on the diffusion couple behavior of U-10Zr (wt%) alloy with Sb against cladding (iron or HT9) at 640°C. The diffusion cross sections were analyzed using a scanning

This study quantifies the uncertainty in the calculated displacements per atom (DPA) value due to uncertainties in the neutron spectrum resulting from cross-section data uncertainty. Using generalized perturbation theory, covariance matrices, and fine-group DPA cross sections, a method for propagating nuclear data uncertainties through to the calculation of DPA is outlined. This method is then implemented

Reynolds-Averaged Navier-Stoke (RANS) models offer an alternative avenue in predicting flow characteristics when the corresponding experiments are difficult to achieve due to geometry complexity, limited budget, or knowledge. RANS models require the knowledge of subgrid scale physics to solve conservation equations for mass, energy, and momentum. Mechanistic turbulence models, such as k-ε, are generally

During the lifetime of fuel assemblies, irradiation and fluid mechanical forces can cause a permanent deformation in the lateral direction that leads to larger interassembly water gaps in the reactor core. The standard reload safety analysis for the reactor core is developed for a uniform distribution of corewise interassembly water gaps. A nonuniform distribution of water gaps with locally larger

Neutron fluences have been measured from 155 MeV/nucleon 4He and 12C ions stopping in an Al target at laboratory angles between 10 and 160 deg. The resultant spectra were integrated over angle and energy above 10 MeV to produce total neutron yields. Comparison of the two systems shows that approximately two times as many neutrons are produced from 155 MeV/nucleon 4He stopping in Al and 155 MeV/nucleon