Abstract A dynamical system under extreme physical disorder has the tendency to evolve toward the equilibrium state characterized by an inverse power law power spectrum. In this paper, a practical, implementable, three-dimensional model is proposed for the random media formed by a multimaterials mixture under such a power spectrum using a randomized form of the Weierstrass function, its extension covering

Abstract The medium-mass structural material titanium has been extensively applied in the nuclear reactor systems of fission or fusion, and related data are also urgently needed. In the present work, all reaction cross sections, angular distributions, energy spectra, and double-differential cross sections are consistently calculated and analyzed for the n+48Ti reaction below 200 MeV. The theoretical

Abstract Jet fragmentation greatly influences the possibility of steam explosion and the formation of a debris bed when a molten corium jet falls into subcooled coolant during a severe accident of a nuclear reactor—which is called fuel and coolant interaction (FCI). The characteristics of different jet fragmentation mechanisms and the conditions under which they play a major role are still in doubt

Abstract A standard approach to solving the S   N transport equations is to use source iteration with diffusion synthetic acceleration (DSA). Although this approach is widely used and effective on many problems, there remain some practical issues with DSA preconditioning, particularly on highly heterogeneous domains. For large-scale parallel simulation, it is critical that both (a) preconditioned source

Abstract Although aerosols in some postaccident nuclear environments can be nonspherical, chainlike, or agglomerates, there have been limited investigations of the rate processes (such as coagulation, evaporation, condensation, and deposition) involving such particles. In a previous investigation, the understandings of condensation and evaporation on such particles were expanded through use of a one-speed

Abstract This paper describes the use principle and application fields of vortex diodes. Unfortunately, all the published studies have not taken into account the restrictions on the coolant flow rate in the vortex diode when used in a nuclear reactor. The diodicity declined significantly due to the limitation of the average flow velocity in the throttle, which does not exceed 12 m/s, while a parallel

Abstract In order to support the development and deployment of uranium fuels with enrichment beyond 5%, additional criticality safety methodologies are needed to prevent the possibility of criticality accidents. Specifically, improved methodologies for computer code validation using evaluated critical experiments, particularly for dissolver systems, need to be developed. Potential candidate evaluations

The molten chloride salt fast reactor (MCFR) with Th-U fuel cycle is attracting more and more attention because of its excellent performance, such as high solubility of actinides, superior breeding capacity, low waste production, and high inherent safety. First of all, the breeding capability of an MCFR at equilibrium state was optimized by an in-house automated optimization program. Based on an optimized

The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production

This work presents an assessment of the applicability of existing benchmark critical experiments to the criticality safety code validation for a large-capacity high-assay low-enriched uranium (HALEU) transportation package concept. Numerous next-generation nuclear reactor designs require HALEU fuel, which is characterized by an enrichment between 5 and 20 wt% 235U. The U.S. Department of Energy (DOE)

The Simplified P N ( S P N ) approximation is often used to model radiation transport phenomena, but it converges to the true solution of the transport equation only in one-dimensional slab geometry. In all other geometries, it incurs a model error that needs to be quantified. In this paper, we estimate the radiation transport model error due to the S P N approximation and employ S N transport solutions

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

Obtaining sufficiently accurate geometric descriptions is a crucial prerequisite for reliable particle transport calculations. Conventional transport algorithms on Cartesian grids use a highly efficient sweep technique and numerous mature discretization methods despite their modeling deficiency for complex geometries. To achieve a more accurate geometric description, a cell-based nonmatching Cartesian

Burnup calculations play a very important role in nuclear reactor design and analysis, and solving burnup equations is an essential topic in burnup calculations. In the last decade, several high-accuracy methods, mainly including the Chebyshev rational approximation method (CRAM), the quadrature-based rational approximation method, the Laguerre polynomial approximation method, and the mini-max polynomial

This work aimed to develop accelerator-driven systems (ADSs) with a subcritical thorium assembly for fuel breeding and clean energy utilization by using several seed fuels. The ADS reactor core was loaded with three different fuel types, namely, reprocessed fuel, UN, and UO2 (seed fuel) associated separately with ThO2 fuel (blanket) in a heterogeneous approach. The Monte Carlo code MCNPX 2.7.0 has

Photon buildup is a function of energy, medium, and geometry and therefore must be specifically calculated for the case of interest. The Martian atmosphere, mostly comprising carbon dioxide, is becoming more relevant to radiation researchers and therefore warrants the study of this gas mixture’s buildup properties for ionizing photon flux resulting from the secondary effects of galactic cosmic rays

Alpha-decay propulsion technology, a microthrust technology based on thin spontaneous-alpha-decay films, is proposed in this paper. A large quantity of decayed alpha particles emitted from the upper surface of thin films would generate thrust statistically. Simulations were executed using the Monte Carlo N-Particle Transport Code (MCNP) to acquire the energy and angular distributions of escaping alpha

This paper is focused on overcoming reactor shutdown malfunction and diagnostics due to poor cooling and water drop level within a pressurized water reactor. So the temperature coupling analysis between the heat exchanger (HEX) and the U-tube steam generator (UTSG) is inspected under changes of primary and secondary water temperature. This coupling allows the removal of heat from the UTSG via the HEX

This paper details and implements a framework for evaluating thermal neutron scattering cross sections that provide S ( α , β ) data and covariance data for hydrogen in light water. This methodology involves perturbing model parameters of molecular dynamics potentials and fitting the simulation results to experimental data. The framework is general and can be applied to any material or simulation method

The void reactivity of a fuel assembly with a streaming channel was measured in a simulated light water reactor critical lattice. The void reactivity was defined as the difference of reactivity ρ between different void conditions. Stainless steel and Zircaloy are candidates for the streaming channel material. Aluminum was used in this measurement because it is inexpensive and its absorption cross section

(2020). Selected Papers from the 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2019). Nuclear Science and Engineering. Ahead of Print.

(2020). Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) Nuclear Science and Engineering: Vol. 194, Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18), pp. 3-4.

(2020). Correction. Nuclear Science and Engineering: Vol. 194, Selected papers from the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18), pp. 1-1.

The gamma irradiation effect in indium phosphide (InP) heterojunction bipolar transistors (HBTs) is studied in this paper. The direct-current (DC) and alternating-current (AC) characteristics are investigated before and after an irradiation dose of 10 Mrad(Si). The main effects of gamma irradiation for InP HBTs are the following: increase of forward Gummel base current at low bias regime, decrease

The dependence of the separative power of Iguassu gas centrifuges (GCs) on the rotor diameter and velocity of rotation V above 1000 m/s is investigated. The separative power is calculated exploring numerical modeling of the gas dynamics and diffusion of the binary mixture in a strong centrifugal field. The separative power is optimized on five internal parameters of the GC: pressure at the wall of

Compression of cross-section data used for high-resolution core analysis is performed using a dimensionality reduction technique based on the singular value decomposition (SVD) and low-rank approximation. The size of cross-section data can be a significant issue in high-resolution core analysis using detailed energy and spatial resolutions. This study addresses this issue focusing on the similarity

Over the years, significant validation work for the neutronics code MPACT has been performed against zero-power critical benchmarks and measured data from operating nuclear power plants. Among all of these efforts, however, validation of the pin-resolved capability in MPACT has been limited by the public availability of experimental data and to a lesser degree availability of measurement techniques

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

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.

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

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 water

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