Abstract This paper presents a new robust scheme for coupled physics nuclear reactor calculations. We focus specifically on high-fidelity whole-core transport calculations with coarse mesh finite difference (CMFD) coupled to thermal hydraulics. These simulations traditionally employ rthe Picard iteration for the coupled solution, where it has been observed that the use of CMFD (or nonlinear diffusion

Abstract Various numerical models are developed that seek to reproduce, in a simulation instance, the formation and evolution of cracks in the claddings of nuclear fuel elements. The algorithms are based on the cohesive zone method within the finite element framework. When applied to simulations involving fracture mechanics, cohesive elements have various advantages, such as not needing to know the

Abstract A benchmark to verify the accuracy of neutron transport criticality solvers along the energy dimension was established. For the first time, the analytic solution of the flux amplitude was derived in the particular case of an infinite-homogeneous medium with isotropic scattering in the center of mass and an arbitrary number of no-threshold, neutral particle reaction resonances (e.g., radiative

Abstract The Parallel Block Jacobi (PBJ) spatial domain decomposition is well suited for implementation on massively parallel computers to solve the neutron transport equation on unstructured grids due to the simple scheduling policy that arises from the PBJ’s iterative asynchronicity. The Parallel Block Jacobi-Integral Transport Matrix Method (PBJ-ITMM) is an iterative method that utilizes the PBJ

Abstract A novel ex-core-detector–based core power reconstruction method is presented. The method uses power correlations between fuel regions and can be applied to a real-time small reactor core monitoring system especially for the detection of abnormal behavior. The use of ex-core detectors reduces the installation and maintenance costs of small modular reactors (SMRs) compared to conventional in-core

Abstract New multiphysics coupling capabilities for molten salt reactor (MSR) analysis have been developed in the Virtual Environment for Reactor Applications (VERA). This development consisted of two main efforts. First, a generic species transport module was added in the CTF code, which is the thermal-hydraulics (TH) code for VERA. This module uses the velocity fields for which CTF solves during

Abstract In this study, a spatio-temporal approach for the solution of the time-dependent Boltzmann (transport) equation is derived. Finding the exact solution using the Boltzmann equation for the general case is generally an open problem and approximate methods are usually used. One of the most common methods is the spherical harmonics method (the PN approximation), when the exact transport equation

Abstract We present a systematic computational dose rate evaluation for a packaged 1.8-Ci 241AmBe source using both Monte Carlo and deterministic approaches, with some experimental measurements for correlation. The 241AmBe source is stored in an extended 55-gal-drum container. Computational dose rate analysis is performed using MCNP6 (Monte Carlo) and PENTRAN (SN) on the Center for High Performance

Abstract The first edition of the current American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard ANSI/ANS-8.1-2014, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,” was published in 1964 as ASA N6.1-1964, “Safety Standard for Operations with Fissionable Materials Outside Reactors.” In 1969, that standard was revised as ANSI N16.1-1969

Abstract Criticality control of random media such as fuel debris is one of the most important safety issues in postaccident management. A 1/fβ spectrum randomizing model is expected to simulate such random media because it is well known that 1/fβ noise can describe a diverse range of random and disordered natural phenomena. In this paper, we focus on the relationship between the multiplication factor

Abstract The Micro-Pocket Fission Detector (MPFD) is a small-form-factor real-time fission chamber. MPFD performance has been simulated in the Advanced Test Reactor Critical Facility located at Idaho National Laboratory. The neutron and gamma-ray flux profiles and magnitudes were simulated using MCNP in the near-core B-8 irradiation position. These simulations were performed at 69 discrete axial locations

Abstract A new dynamic verification of the one-dimensional (1-D) computational Two-Fluid Model (TFM) using the Type II density wave instability (DWI) theory of Ishii is presented. Verification requires convergence in the sense of the Lax Equivalence Theorem and dynamic comparison with the DWI theory. Rigorous verification of the computational TFM must be performed with a computational model that is

Abstract In the face of advanced persistent threat actors, existing information technology (IT) defenses as well as some of the more recent operational technology (OT) defenses have been shown to become increasingly vulnerable, especially for critical infrastructure systems with well-established technical know-how. For example, data deception attacks have demonstrated their ability to mislead human

Abstract In order to ensure safety in a nuclear power plant, operation and protection systems must take into account safety parameters, whether to guide operators or to trip the reactor in emergency cases. Especially in a boron-free small modular reactor (SMR) where reactivity and power are controlled exclusively by rod banks, the power distribution is mostly influenced by its movements affecting the

Abstract The optimization of measurement of environmental samples is achieved by putting the sample closer to the detector to increase the full-energy peak efficiency, which leads to decrease of the detection limit. The present work inspects the utilization of Geant4 simulation for a NaI cubic scintillation detector with a cavity using two tracks. The radionuclide option includes coincidence summing

Abstract In this paper, detailed verification and experimental validation of the formulations and algorithms of the Multi-stage Response-function Transport (MRT)–based Real-time Analysis for Particle-transport and In-situ Detection (RAPID) code system is presented. In particular, RAPID’s fission matrix formulation for eigenvalue calculations and its detector response function for reaction rate calculations

Abstract The deployment of molten salt reactors requires validation of the computational tools used to support the licensing process. The Molten Salt Reactor Experiment (MSRE), built and operated in the 1960s, offers a unique inventory of experimental data for reactor physics benchmarks. The first benchmark based on the MSRE appeared in “The 2019 Edition of the IRPhEP [International Reactor Physics

Abstract Computing in the energy dimension is one of the greatest challenges confronting present-day deterministic neutron transport solvers. Accurately resolving the neutron flux as neutrons downscatter across resonances in the nuclear cross sections currently requires considerable computing power and suffers from approximation errors. Flux uncertainty resulting from the uncertainty of the resonance

Abstract For full-core modeling in the Virtual Environment for Reactor Analysis (VERA), the three-dimensional multigroup eigenvalue neutron transport problem is solved by MPACT. To improve the efficiency of MPACT, advancements have been made in the transport accelerator. Multilevel-in-energy and multilevel-in-space coarse mesh finite difference (CMFD) solvers were developed to improve the efficiency

Abstract A target velocity sampling method named the Relative Speed Tabulation (RST) is proposed for the efficient treatment of resonance elastic scattering in the Monte Carlo simulation utilizing graphics processing units (GPU). The RST method samples the relative speed between a neutron and a target nucleus by employing pretabulated probabilities of relative speeds. The target velocity is then determined

Abstract Aiming at studying the condensate flow phenomenon and air-steam–mixture condensation heat transfer underneath a containment vessel surface, a test bench was constructed. The plate dimension was 1.5 × 0.6 m, with Carbozinc 11 coating on the surface, suspended in a pressure vessel with 2.5-m diameter and 4.5-m height. The air-steam mixture was condensed on an inclined plate through natural convection

Abstract Cumulative and independent yields of various fission products within mass ranges of 86 to 119 and 127 to 162 were measured in the spontaneous fission of 252Cf by using an off-line gamma-ray spectrometric technique. From the cumulative yields of the fission products, their mass chain yields were obtained by using the charge distribution correction. Mass yield distribution parameters such as

Abstract Neutron capture and transmission measurements were carried out from thermal to 2000 eV on both solid and liquid samples containing elemental cesium (133Cs). This work describes the extension of the R-matrix analysis of these data from 600 to 2000 eV by correcting the capture data for false capture in the NaI detector. These false capture–corrected capture and transmission data were analyzed

Abstract In this paper, we investigate the possibility of plutonium quantification within the electrorefiner vessel of an electrochemical separation facility via the use of the (α,n) neutron signature from dissolved actinides. As a potential alternative means to traditional spontaneous fission tracking, such an analysis may provide a more reliable tracking capability of plutonium within systems that

Abstract The Transient Multilevel (TML) scheme in the MPACT code has reduced the computational burden for three-dimensional, full-core, time-dependent reactor simulations with pin-resolved detail. However, the total computational cost is still large for practical applications. In this paper, we present a new method that uses a one-group coarse mesh finite difference (1GCMFD) approach to further accelerate

Abstract Uncertainty quantification of lead (Pb) and bismuth (Bi) sample reactivity worth is numerically determined using the SCALE6.2 code system and experimental results obtained from the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA) to demonstrate the sensitivity coefficients of aluminum (Al) and Bi scattering reactions. From the results of the numerical

Abstract Recent efforts in MPACT have focused on improving the performance of the 2D/1D subplane implementation to help target computational performance goals. This paper builds on previous efforts that targeted the use of subgrid treatments to improve the accuracy of control rod representation, presenting three additional applications of subgrid treatments with the goal of reducing the computational

Abstract Based on embedded analysis, an accurate pin power reconstruction (PPR) method is proposed for conventional nodal analysis. Unlike the common form function (FF) method, the new PPR method, named the embedded pin power reconstruction (EPPR) method, directly solves a two-group fixed-source problem that is defined with pinwise homogenized group constants (HGCs) and coarse-mesh incoming partial

Abstract The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas flows from the drywell to the wetwell after the normally submerged connecting vents between the drywell and the wetwell have been purged of water through a vent clearing process so that the suppression

(2021). Correction. Nuclear Science and Engineering: Vol. 195, No. 3, pp. 338-338.

Abstract Concrete plays a major role in nuclear facilities as protection against radiation. However, its chemical composition, which is fundamental, is often unknown. Several concrete samples, extracted from the AMANDE-MIRCOM Institute for Radiological Protection and Nuclear Safety (IRSN) facility, were analyzed. Various simulations were performed in order to evaluate the neutron fluence behind a 40-cm-thick

Abstract Concrete plays a major role in nuclear facilities as protection against radiation. However, its chemical composition, which is fundamental, is often unknown. Several concrete samples, extracted from the AMANDE-MIRCOM Institute for Radiological Protection and Nuclear Safety (IRSN) facility, were analyzed. Various simulations were performed in order to evaluate the neutron fluence behind a 40-cm-thick

Abstract In order to ensure safety in a nuclear power plant, operation and protection systems must take into account safety parameters, whether to guide operators or to trip the reactor in emergency cases. Especially in a boron-free small modular reactor (SMR) where reactivity and power are controlled exclusively by rod banks, the power distribution is mostly influenced by its movements affecting the

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 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 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 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 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 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 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 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 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 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 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

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, ...

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 hav...

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