Abstract To determine the angular discretization error asymptotic convergence rate of the uncollided scalar flux computed with the Discrete Ordinates (S   N ) method, a comprehensive theory of the regularity order with respect to the azimuthal angle of the exact pointwise SN uncollided angular flux is derived based on the integral form of the transport equation in two-dimensional Cartesian geometry

Abstract A Resonance calculation using energy Spectrum Expansion (RSE) method is newly proposed in this paper. In this method, ultra-fine-group (UFG) spectra appearing in a resonance calculation are expanded by orthogonal bases on energy, which are extracted from the UFG spectra obtained in homogeneous geometry with various background cross sections using singular value decomposition and low-rank approximation

Abstract The analysis of the results of a depletion code is often considered a tedious and delicate task, for it requires both the processing of large volumes of information (the time-dependent composition of up to thousands of isotopes) and an extensive knowledge of nuclear reactions and associated nuclear data. From these observations, dedicated developments have been integrated to the upcoming version

Abstract Traditional light water reactor simulations are usually either high fidelity, requiring hundreds of node-hours, or low fidelity, requiring only seconds to run on a common workstation. In current research, it is desirable to combine the positive aspects of both of these simulation types while minimizing their associated negative costs. Because neural networks have shown significant success

Abstract In this paper, a beta radioluminescent battery with different radioisotopes is studied, and different parameters of the proposed structure are optimized. These parameters include the luminescent layer thickness, the doping concentration in the semiconductor P-N junction, etc. Some of the parameters have an inverse effect on the battery outputs. So, a trade-off is sought between them to increase

Abstract Growth of a nonisothermal bubble on a heated horizontal surface in a subcooled flow is studied to determine the significance of different heat transfer mechanisms on the bubble growth. The heat transfer mechanisms that are considered are (1) microlayer evaporation, (2) transient thermal boundary layer conduction, and (3) bubble surface evaporation and condensation. The results indicate that

Abstract The Gattar area is classified as one of the most important Egyptian uranium resources. In this work, the leaching of uranium from Gattar granitic ore containing 1138 ppm U using NaNO3 solution has been applied. The leaching factors of nitrate in the presence and absence of hydrogen peroxide as oxidant were studied and optimized. The obtained data show that uranium leaching in the absence of

Abstract Yttrium hydride is being considered as a moderator material for microreactor concepts because of its excellent hydrogen retainment capacity at high temperatures. These types of reactors, operating at thermal to epithermal neutron energies, require accurate thermal scattering laws (TSLs) for yttrium hydride to predict and optimize moderator performance. Currently, TSL evaluations exist only

Abstract To proceed with the decommissioning of the Fukushima Daiichi Nuclear Power Station, analyses of unexpected fuel debris criticality accidents are needed. Supercritical transient analyses have been conducted for fuel debris using the Multiregion Integral Kinetic (MIK) code, which can take the space dependence of fuel debris into account. In those analyses, reactivity is assumed as stepwise insertion

Abstract A thermomechanical fuel performance analysis module is implemented in the Korea Advanced Institute of Science and Technology Monte Carlo (MC) neutron transport code iMC. The module is designed particularly for advanced three-dimensional (3-D) fuel concepts, so an unstructured tetrahedral mesh grid is adopted for geometry flexibility. The cellwise detailed power density distribution is tallied

Abstract The complex structure of a pool-type sodium-cooled fast reactor may lead to uncertainty and asymmetry of flow and temperature field distributions under a pump stuck accident. This phenomenon has obvious three-dimensional (3-D) thermal-hydraulic characteristics and cannot be analyzed by one-dimensional or two-dimensional models. Previous research has been limited and lacking of 3-D numerical

Abstract The System Analysis Module (SAM), developed and maintained by Argonne National Laboratory, is designed to provide whole-plant transient safety analysis capabilities for a number of advanced non–light water reactors, including sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and molten salt reactor (MSR)/fluoride-salt-cooled high-temperature reactor (FHR) designs. SAM is primarily

Abstract The neutron spectrum is an essential factor in making possible the increase of 233U isotope breeding from thorium fuel in an accelerator-driven subcritical (ADS) system; therefore, studying the effects of various moderators and coolants on 233U breeding is an important step in ADS performance. This study aims to evaluate the effect of using different moderators and coolants on the ADS system

The reviewers’ identities are traditionally kept anonymous in peer review, yet their dedication to the arduous task of reviewing articles for Nuclear Science and Engineering is indispensable to the validation and dissemination of valuable scientific information. By listing the reviewers, I am expressing both my personal gratitude and the appreciation of the authors for the reviewers’ invaluable contributions

Abstract Recently completed cross-section evaluations sponsored in part by the Nuclear Criticality Safety Program were incorporated into the 2018 release of the ENDF/B-VIII.0 cross-section library. Evaluated isotopes of interest to the nuclear data and criticality safety community include 16O, 56Fe, and 63,65Cu. For performance validation, benchmark models defined in the International Criticality Safety

Abstract While the literature has numerous examples of Monte Carlo (MC) and computational fluid dynamics (CFD) coupling, most are hardwired codes intended primarily for research rather than as stand-alone, general-purpose applications. In this work, we describe an open source application, the Exascale Nuclear Reactor Investigative COde (ENRICO), which enables coupled neutronic and thermal-hydraulic

Abstract We developed a cross-section adjustment method based on Bayesian theory for adjusting the independent cross-section set that is given by dividing the cross-section set. While all cross sections have to be adjusted in the conventional cross-section adjustment method, the independent cross-section set can be individually adjusted in the developed method. In addition, the covariance of cross

Abstract The CASL reactor simulation package VERA has been adapted to provide high-fidelity simulation capabilities for modeling source range detector response during subcritical reactor configurations. New features include the activation and shuffling of secondary-source assemblies, use of burned fuel neutron emission data from the ORIGEN depletion solver to the MPACT deterministic neutron transport

Abstract In order to obtain the heat transfer characteristics of the helium gap in conditions of different material thicknesses and linear power in high-temperature ranges, on the basis of previous research, an existent test device was improved. Through the theoretical design of double helium, the test device can perform experiments under high-temperature conditions. Compared with the experimental

Abstract Computational modeling and simulations are widely used for evaluation of the performance and safety features of innovative nuclear reactor designs. Multigroup-based deterministic neutronics codes are often employed in these reactor design calculations because they can provide fast predictions of the neutron flux distribution and other neutronics characteristic parameters. Nevertheless, providing

Abstract Scattering of thermal neutrons and Doppler broadening of epithermal neutron resonances in uranium and its compounds may be sensitive to crystal binding. The thermal scattering law (TSL) for uranium dioxide, which captures crystal binding effects, has been reevaluated for ENDF/B-VIII.0. Phonon spectra were generated using ab initio lattice dynamics for the paramagnetic phase and validated against

Abstract This work presents a Reynolds-averaged Navier Stokes–based computational fluid dynamics methodology for the calculation of pressure drop and mass flow rate distribution in a material test reactor flat-plate-type standard fuel assembly (SFA) of the IEA-R1 Brazilian research reactor to predict future improvements in newer SFA designs. The results improve the understanding of the origin of fuel

(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: Vol. 194, Selected papers from the 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2019), pp. iii-iv.

Abstract Fluctuations associated with power and detector readings in a nuclear reactor, commonly known as reactor noise, are of great importance in nuclear science and engineering. Two different types of noise are described in the literature: internal noise, which is associated with the inherent stochasticity of fission chains, and external noise, which is governed by physical fluctuations of the macroscopic

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 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

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

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

Abstract 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

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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

(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.

Abstract 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,

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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

Abstract 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