(2021). Special issue on the National Criticality Experiments Research Center. Nuclear Science and Engineering: Vol. 195, No. sup1, pp. iii-iv.

(2021). Editor’s Note. Nuclear Science and Engineering: Vol. 195, No. sup1, pp. v-v.

Abstract Planet is a vertical-lift assembly machine currently located at the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site. In the past, Planet resided at Technical Area-18 in Los Alamos, New Mexico, as part of the Los Alamos Critical Experiments Facility (LACEF). Following the de-inventorying of LACEF, the Planet assembly was relocated to NCERC in 2008

Abstract Comet is a general-purpose, heavy-duty, vertical-lift assembly designed for flexibility in conducting a variety of critical experiments. It is currently located at the National Criticality Experiments Research Center (NCERC) in Nevada. In the past, Comet resided at Technical Area-18 in Los Alamos, New Mexico, as part of the Los Alamos Critical Experiments Facility (LACEF). The Comet assembly

Abstract The Flattop critical assembly was built and operated from 1958 through 2004 at the Pajarito Site (Technical Area-18), home to the Los Alamos Critical Experiments Facility. Flattop was disassembled in 2005 and refurbished over the course of 3 years. In 2008, Flattop was installed in the National Criticality Experiments Research Center (NCERC) located in the Device Assembly Facility at the Nevada

Abstract The work presented in this paper focuses on the first 10 years (2011–2020) of Godiva IV operations at the National Criticality Experiments Research Center (NCERC). Godiva IV is a fast burst critical assembly constructed of approximately 65 kg of highly enriched uranium fuel alloyed with 1.5% molybdenum for strength. Godiva is one of the last such critical assemblies in the United States and

Abstract The work presented in this paper focuses on the first 10 years (2011–2020) of radiation test object (RTO) operations at the National Criticality Experiments Research Center. RTOs are subcritical configurations of special nuclear material that are built by hand. These configurations are utilized for benchmark experiments, detector testing/characterization, and training. An overview of the types

Abstract Nuclear power pipeline ice plug technology, as the main technical method of nuclear power safety maintenance and innovative design, plays a protective role in the long-term safe operation of nuclear power plants. During the ice plug process, pipelines have complex stress states, which can easily lead to problems such as material yielding, cracking, and even damage. In this study, by building

Abstract To simulate the complex accident phenomena of a marine reactor, the thermal-hydraulic system code RELAP5 is modified to perform the analysis under ocean conditions. An integrated reactor with a passive residual heat removal system (PRHRS) is modeled by the improved code, and the effects of different ocean motions under a total loss-of-flow accident (LOFA) and a loss-of-heat-sink (LOHS) accident

Abstract In a previous paper by Pázsit and Pál [“Multiplicity Theory Beyond the Point Model,” Ann. Nucl. Energy, Vol. 154 (2021)], a general transport theory calculation of the factorial moments of the number of neutrons emitted spontaneously from a sample was elaborated. In contrast to the original derivations by Hage and Cifarelli [“On the Factorial Moments of the Neutron Multiplicity Distribution

(2021). ADDENDUM: Validation of Continuous-Energy ENDF/B-VIII.0 16O, 56Fe, and 63,65Cu Cross Sections for Nuclear Criticality Safety Applications. Nuclear Science and Engineering. Ahead of Print.

Abstract Since nuclear power plants were first developed over 80 years ago, most have had thermal reactors with enriched uranium as the main fuel. As a result, huge amounts of depleted uranium must be stockpiled and stored. Fast reactors operating in breed-and-burn (B&B) conditions are one potential approach to reducing the need for storage facilities to hold spent nuclear fuel and maximize the effective

Abstract This study focuses on the calculation of the energy deposition in the Transformational Challenge Reactor by two major Monte Carlo codes: Serpent and MCNP. The first software computation relies on Kinetic Energy Released per unit Mass (KERMA) factors while the second one relies on Q-values. The results from these two independent computation methodologies are in very good agreement; however

Abstract The sodium flow resistance in sodium-cooled fast reactor cores experiencing natural circulation conditions was measured for wire-wrapped 19- and 37-pin bundles using low-velocity water flows with Re <1000 and Re <750, respectively. The measurements were compared with predictions of existing wire-wrapped bundle friction factor correlations. The results show that the existing correlations usually

Abstract Measurements of 243Am and 235U fission reaction rates are conducted with the use of two single fission chambers in the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA). Critical irradiation experiments of 243Am and 235U foils are carried out, and the measured result of 243Am/235U is 0.0424 ± 0.0019; also, calculation/experiment values between calculated

Abstract The power manuverability of nuclear power plants (NPPs) is becoming more and more important as governments expand the deployment of renewable energy resources in their energy mix. For this purpose, load follow (LF) operation (LFO) schemes are introduced and tested for different types of reactors. Currently, all NPPs in Korea are operated at a baseload, that is, 100% rated power, and do not

Abstract To robustly estimate the fundamental mode component of prompt neutron decay constant α in a subcritical system, dynamic mode decomposition (DMD) is applied to time-series data obtained by the pulsed-neutron source (PNS) and Rossi-α methods. For the statistical uncertainty quantification of α by DMD, randomly sampled virtual data are used for the DMD procedure. The applicability of DMD is demonstrated

Abstract In a high dominance ratio system, the problem of slow fission source convergence is faced during the Monte Carlo criticality calculation. The Wielandt method and the Superhistory method have been proven to reduce the dominance ratio. Still, the Wielandt method and the Superhistory method have also proven to be unable to accelerate the convergence of the fission source. With the estimation

Abstract Methods for approximately accounting for the terms neglected in a finite (L’th-order) Legendre expansion of the scattering source in the transport equation are called transport corrections. This paper derives adjoint-based sensitivities of a neutron or gamma-ray transport response for problems that use diagonal, Bell-Hansen-Sandmeier (BHS), or n’th-Cesàro-mean-of-order-2 (Cesàro) transport

Abstract Three new rejection sampling methods for generating samples from the adjoint Klein-Nishina cross section are discussed: the two-branch rejection sampling procedure, the three-branch linear rejection sampling procedure and the three-branch inverse rejection sampling procedure. These methods have all been implemented in the Framework for REsearch in Nuclear ScIence and Engineering (FRENSIE)

Abstract Neutron scattering from a copper sample was measured at Rensselaer Polytechnic Institute utilizing the quasi-differential method. The measurement spanned the energy range from 0.5 to 20 MeV using the high-energy scattering system and from 2 keV to 0.5 MeV using the new mid-energy scattering system. Copper was selected as a material of interest to measure due to large discrepancies between

Abstract Mass yield distribution in the epi-cadmium neutron-induced fission of 237Np has been carried out by measuring the cumulative yields of fission products within the mass ranges of 78 to 117 and 123 to 157. A radiochemical and off-line gamma-ray spectrometric technique was used for the measurement. From the cumulative yields of the fission products, mass chain yields were obtained by applying

Abstract Monte Carlo methods are useful for simulating new reactor designs, but even with advances in computing, these methods still require a significant amount of time to perform transient or multiphysics calculations coupled with thermal modeling. This work demonstrates a hybrid reactor physics method that uses Monte Carlo to precalculate an initial database of fission matrix parameters, then combines

Abstract This work introduces a meshed debonding model intended for use with tristructural isotropic (TRISO) fuel particles. The ability to better examine the potential effects of interface debonding between the coating layers will enable a host of investigations of the stress state when complex loading and interactions between coated fuel particles are taking place. In this work, the debonding model

Abstract Energy discretization of the transport equation is difficult due to numerous strong, narrow cross-section (XS) resonances. The standard traditional multigroup (MG) method can be sensitive to approximations in the weighting spectrum chosen for XS averaging, which can lead to inaccurate treatment of important phenomena such as self-shielding. We generalize the concept of a group to a discontiguous

Abstract The multidual differentiation method has been implemented in a ray-tracing transport simulation for the purpose of calculating arbitrary-order sensitivities of the uncollided particle leakage. This method extends dual number differentiation by perturbing variables along multiple nonreal axes to calculate arbitrary-order derivatives. Numerical results of first-through third-order multidual

Abstract Conventionally, the stability of xenon oscillations is estimated by solution of the time-dependent neutron diffusion equation, coupled with iodine and xenon equations, by finding out the damping ratios in each case. This is performed for different initial perturbations and core burnup conditions and is a very time-consuming and tedious process. Some earlier studies include linear stability

Abstract The postulated dual-failure accident, i.e., loss of primary coolant flow along with impairment of the emergency coolant injection system, leads to peak fuel temperatures. It is well known that the temperature of the fuel assemblies is one of the significant factors that affect the outcome of an accident. Therefore, the present work aims to thoroughly investigate the thermal response of a single

Abstract New fuel design and development currently require 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced-scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance

Abstract The neutron transport equation is usually solved against a stationary background medium. When the background material is moving, the transport equation will need to be modified. Solving the transport equation with moving material is quite complicated, especially for the curved coordinate system because of the double angular redistributions. In this study, the discretization method of the simplified

Abstract An optical model potential for neutron-antimony isotopes collision systems is obtained at the incident energies up to 200 MeV. Cross sections, angular distributions, energy spectra and double differential cross sections for neutron-induced reactions on 121Sb, 123Sb and natural Sb are consistently calculated and analyzed at incident energies below 20 MeV by using the theoretical nuclear models

Abstract The objective of the present study is to show that it is feasible to establish the breed-and-burn (B&B) mode of operation with rotational fuel shuffling in the S-PRISM core based on neutronic and thermal-hydraulic analyses. The results quantified the impact of major core design choices on the criticality of a core that uses sodium as a coolant material and HT9 steel as structural material

Abstract A variety of normal operation and accident scenarios can generate thermal stresses large enough to cause cracking in ceramic fuel pellets. Cracking in fuel pellets can lead to reduced heat removal, higher centerline temperatures, and localized stress in the cladding—all which impact fuel performance. It is important to experimentally characterize the thermal and mechanical behaviors in the

Abstract It has been well known that the analytic neutron transport solution tends to the analytic solution of a diffusion problem for optically thick systems with small absorption and source. The standard technique for proving the asymptotic diffusion limit is constructing an asymptotic power series of the neutron angular flux in small positive parameter ε, which is the ratio of a typical mean free

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 Algorithms used to generate Monte Carlo input decks and to analyze the output over a range of uranium mass, water volume, and particle size in a regular lattice are described. The algorithms produce input decks for both homogeneous and heterogeneous, regular-lattice systems of 20% enriched uranium metal and water and then analyze the results to determine the minimum critical mass over a range

Abstract In this work, we aim to show that machine learning algorithms are promising tools for the identification of nuclear data that contribute to increased errors in transport simulations. We demonstrate this through an application of a machine learning algorithm (Random Forest) to the Whisper/MCNP6 criticality validation library to identify nuclear data that are associated with an increase of the

Abstract Because of the complexity of the nuclear reactor system, traditional statistical sampling methods, such as random sampling and Latin hypercube sampling, often lead to unstable uncertainty quantification results of the reactor physics analysis. In order to make the analysis results robust, traditional sampling methods require a large number of samples, which brings a huge computation cost.

Abstract This study presents the Probabilistic Seismic Demand Model (PSDM) and explores optimal intensity measures (IMs) for nuclear power plant (NPP) equipment when subjected to ground motions having high-frequency (HF) and low-frequency (LF) contents. To this end, the PSDM is first constructed in terms of the IM and the corresponding engineering demand parameter, and its outcomes are utilized to

Abstract This study introduces new methodologies for integrating fission reactions induced by delayed neutrons into the Multi-Region Integral Kinetic (MIK) code by using a Monte Carlo neutron transport calculation. First, it was confirmed that it is feasible to solve the Integral Kinetic Model (IKM) with delayed neutrons by the forward Euler discretization method in terms of the number of time steps

Abstract Using time-of-flight techniques, we have measured neutron emission spectra from 6- and 10-MeV incident neutron energies on 10B and 11B for laboratory angles between 20 and 145 deg using the unique neutron source reaction 1H(t,n). Double-differential cross sections and their integrated values have been extracted for elastic and inelastic processes, and their integrated values are summarized

Abstract In a companion paper, we present the theoretical development of a new robust, relaxation-free iteration scheme for multiphysics k-eigenvalue problems. These types of problems are essential to the study of computational reactor physics and in particular whole-core, high-fidelity simulation codes. The deterministic whole-core simulation tools invariably rely on the coarse mesh finite difference

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 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 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 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