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At least 1,747 records · Page 97

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DOE OSTI2022

Unstructured mesh-based multi-physics reactor transient simulation with iMC code

In this study, we present a time-dependent Monte Carlo simulation result about a complex fuel-loaded light water reactor system with thermal-hydraulic feedback using the KAIST iMC code. The intra-pin temperature distribution is calculated based on Monte Carlo-tallied detailed power distribution and the finite element heat transfer method. The temperature effect of fuel, absorber, and coolant are adequately reflected into the dynamic response by the on-the-fly Doppler broadened cross-section correction. For the time-dependent Monte Carlo transport simulation, the predictor-corrector quasi-static Monte Carlo method is used for this study. We tested the coupled analysis framework with a 17-by-17 fuel assembly loaded with the centrally-shielded burnable absorber (CSBA) fuel element. The temperature feedback effect on the material cross-section and the moderator density incurred negative reactivity in response to the external positive reactivity insertion, showing the inherent stability of the reactor system. The pin-wise power and temperature change during the system transient are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Electron Accelerator Shielding Design of KIPT Neutron Source Facility

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, similar to 0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper. Copyright (C) 2016, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.

43 PARTICLE ACCELERATORS↗

NASA NTRS2022

Numerical Results for Polarized Light Scattering in a Spherical Atmosphere

We report numerical results for polarized light reflection from the top of a Rayleigh scattering spherical atmosphere with height-dependent single scattering albedo over a dark surface. Michael Mishchenko considered this scenario back in the 1990’s, for a plane-parallel atmosphere of unit optical thickness (OT), for which radiance errors arising from neglecting polarization reaches their highest values. To further extend Mishchenko’s results, we consider a value of OT = 0.25, for which the effect of atmospheric curvature is pronounced. New results are generated using three state-of-the art radiative transfer (RT) codes. These are: the MYSTIC and MCSSA models, which simulate light scattering in a true- spherical atmosphere using Monte Carlo methods; and the discrete ordinate code VLIDORT, operating with a new multiple-scatter spherical correction designed to deliver reasonable approximations to spherical-medium scattering. In this work, we report results for both single and multiple scattering; this will help to support the validation of existing and future polarized spherical RT codes, especially those using approximative methods to deal with sphericity.

S Korkin↗

DOE OSTI2012

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Hexagonal Uniformly Redundant Arrays (HURAs) for scintillator based coded aperture neutron imaging

A series of Monte Carlo simulations have been conducted, making use of the EJ-426 neutron scintillator detector, to investigate the potential of using hexagonal uniformly redundant arrays (HURAs) for scintillator based coded aperture neutron imaging. This type of scintillator material has a low sensitivity to gamma rays, therefore, is of particular use in a system with a source that emits both neutrons and gamma rays. The simulations used an AmBe source, neutron images have been produced using different coded-aperture materials (boron- 10, cadmium-113 and gadolinium-157) and location error has also been estimated. In each case the neutron image clearly shows the location of the source with a relatively small location error. Neutron images with high resolution can be easily used to identify and locate nuclear materials precisely in nuclear security and nuclear decommissioning applications. (authors)

42 ENGINEERING↗

NASA NTRS1972

Probable solar flare doses encountered on an interplanetary mission as calculated by the MCFLARE code

The computer program, MCFLARE, uses Monte Carlo methods to simulate solar flare occurrences during an interplanetary space voyage. The total biological dose inside a shielded crew compartment due to the flares encountered during the voyage is determined. The computer program evaluates the doses obtained on a large number of trips having identical trajectories. From these results, a dose D sub p having a probability p of not being exceeded during the voyage can be determined as a function of p for any shield material configuration. To illustrate the use of the code, a trip to Mars and return is calculated, and estimated doses behind several thicknesses of aluminum shield and water shield are presented.

Lahti, G. P.↗

DOE OSTI2015

Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (k eff ) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of k eff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δk eff (ISG-8, Rev. 3, Recommendation 4).

42 ENGINEERING↗

DOE OSTI2014

Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Radiation Source Localization Using Surrogate Models Constructed from 3-D Monte Carlo Transport Physics Simulations

Recent research has focused on the development of surrogate models for radiation source localization in a simulated urban domain. We employ the Monte Carlo N-Particle (MCNP) code to provide high- delity simulations of radiation transport within an urban domain. The model is constructed to employ a source location (x, y, z) as input and return the estimated count rate for a set of speci ed detector locations. Because MCNP simulations are computationally expensive, we develop e cient and accurate surrogate models of the detector responses. We construct surrogate models using Gaussian processes (GP) and neural networks (NN) that we train and verify using the MCNP simulations. The trained surrogate models provide an e cient framework for Bayesian inference and experimental design. We employ Delayed Rejection Adaptive Metropolis (DRAM), a Markov Chain Monte Carlo (MCMC) algorithm, to infer the location and intensity of an unknown source. The DRAM results yield a posterior probability distribution for the source's location conditioned on the observed detector count rates. The posterior distribution exhibits regions of high and low probability within the simulated environment identifying potential source locations. In this manner, we can quantify the source location to within at least one of these regions of high probability in the considered cases. Employing these methods, we are able to reduce the space of potential source locations by at least 60%.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2014

Numerical simulation of ion charge breeding in electron beam ion source

The Electron Beam Ion Source particle-in-cell code (EBIS-PIC) tracks ions in an EBIS electron beam while updating electric potential self-consistently and atomic processes by the Monte Carlo method. Recent improvements to the code are reported in this paper. The ionization module has been improved by using experimental ionization energies and shell effects. The acceptance of injected ions and the emittance of extracted ion beam are calculated by extending EBIS-PIC to the beam line transport region. An EBIS-PIC simulation is performed for a Cs charge-breeding experiment at BNL. The charge state distribution agrees well with experiments, and additional simulation results of radial profiles and velocity space distributions of the trapped ions are presented.

43 PARTICLE ACCELERATORS↗

DOE OSTI2019

Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset, known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. Depletion capability is supported by coupling SCALE-ORIGEN with MPACT. A new ENDF/B-VII.1 MPACT 51-group library has been developed by using the AMPX/SCALE and CASL-XSTools code packages based on the procedure introduced in the companion paper, “Development of the Multigroup Cross Section Library for the CASL Neutronics Simulator MPACT: Method and Procedure.” The cross section library generation procedure, cross section library, and MPACT transport capability have been verified by performing various benchmark calculations and comparing the benchmark results to the continuous-energy (CE) Monte Carlo results. Furthermore, the code-to-code comparison for the benchmarks shows that the CASL neutronics simulator MPACT and its 51-group library are working reasonably.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2005

Unified Models of Turbulence and Nonlinear Wave Evolution in the Extended Solar Corona and Solar Wind

The paper discusses the following: 1. No-cost Extension. The no-cost extension is required to complete the work on the unified model codes (both hydrodynamic and kinetic Monte Carlo) as described in the initial proposal and previous annual reports. 2. Scientific Accomplishments during the Report Period. We completed a comprehensive model of Alfvtn wave reflection that spans the full distance from the photosphere to the distant heliosphere. 3. Comparison of Accomplishments with Proposed Goals. The proposal contained two specific objectives for Year 3: (1) to complete the unified model code, and (2) to apply it to various kinds of coronal holes (and polar plumes within coronal holes). Although the anticipated route toward these two final goals has changed (see accomplishments 2a and 2b above), they remain the major milestones for the extended period of performance. Accomplishments la and IC were necessary prerequisites for the derivation of "physically relevant transport and mode-coupling terms" for the unified model codes (as stated in the proposal Year 3 goals). We have fulfilled the proposed "core work" to study 4 general types of physical processes; in previous years we studied turbulence, mode coupling (Le., non-WKB reflection), and kinetic wave damping, and accomplishment lb provides the fourth topic: nonlinear steepening.

Wagner, William↗

DOE OSTI2012

Uncertainty Quantification with Monte Carlo Hauser-Feshbach Calculations of Prompt Fission Neutrons and Gamma Rays

We have performed sensitivity calculations using the CGMF code that implements the statistical Hauser-Feshbach equations in a Monte Carlo fashion to describe the de-excitation of fission fragments by emis sion of prompt neutrons and gamma rays. The specific importance of several input quantities is emphasized. An uncertainty quantification method, considering sources of uncertainty from both model calculations and experiments, is outlined.

Atomic and Nuclear Physics↗

DOE OSTI2012

MCNP6 for Criticality Accident Alarm Systems – A Primer

MCNP6 is a Los Alamos National Laboratory (LANL) developed, general-purpose Monte Carlo radiation transport simulation package. The code is capable of transporting a broad range of particles for numerous applications. MCNP6 has continuous-energy neutron and photon physics that make it suitable for both criticality and shielding problems. The analysis of a criticality accident alarm system (CAAS) combines methodologies and techniques for both classes of problems, and the requisite knowledge for both is quite different in some areas. For this reason, a primer has been created with the focus on teaching criticality safety practitioners the Monte Carlo techniques needed to analyze a CAAS. This primer specifically focuses on the Monte Carlo transport techniques for CAAS design. The use of radiation transport codes is a small portion of the overall design process, which is preceded and followed by many other design calculations and decisions. Those readers unfamiliar with the design of CAAS systems are directed to a PNNL report by Greenfield, which gives a thorough overview of the CAAS design process. This document gives reference to applicable standards and regulatory requirements and how to meet them, which is not the subject of this primer. The PNNL report explains where radiation transport codes such as MCNP play a role in this process, and this primer is meant to provide a more in-depth set of lessons for this portion of the design process.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2016

Optimal Temperature Grid for Accurate Doppler Kernel Reconstruction

For computational efficiency, both Monte Carlo and deterministic neutron transport codes often rely on pre-tabulated cross section sets and simple interpolation schemes to represent the Doppler broadening of the cross sections at a needed temperature. Accurate accounting of Doppler effects is of paramount importance for reactor analysis since it impacts the power distribution and the reactor response during transients. Temperature interpolation methods for Doppler broadening have traditionally been physically motivated engineering curve-fits, selected upon best-performance criteria on important nuclides.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Evidence for deconfined U(1) gauge theory at the transition between toric code and double semion

Building on quantum Monte Carlo simulations, we study the phase diagram of a one-parameter Hamiltonian interpolating between trivial and topological Ising paramagnets in two dimensions, which are dual to the toric code and the double semion. We discover an intermediate phase with stripe order which spontaneously breaks the protecting Ising symmetry. Remarkably, we find evidence that this intervening phase is gapless due to the incommensurability of the stripe pattern and that it is dual to a $U(1)$ gauge theory exhibiting Cantor deconfinement.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2016

MC3, Version 1

The MC3 code is used to perform Monte Carlo simulations in the isothermal-isobaric ensemble (constant number of particles, temperature, and pressure) on molecular crystals. The molecules within the periodic simulation cell are treated as rigid bodies, alleviating the requirement for a complex interatomic potential. Intermolecular interactions are described using generic, atom-centered pair potentials whose parameterization is taken from the literature [D. E. Williams, J. Comput. Chem., 22, 1154 (2001)] and electrostatic interactions arising from atom-centered, fixed, point partial charges. The primary uses of the MC3 code are the computation of i) the temperature and pressure dependence of lattice parameters and thermal expansion coefficients, ii) tensors of elastic constants and compliances via the Parrinello and Rahman’s fluctuation formula [M. Parrinello and A. Rahman, J. Chem. Phys., 76, 2662 (1982)], and iii) the investigation of polymorphic phase transformations. The MC3 code is written in Fortran90 and requires LAPACK and BLAS linear algebra libraries to be linked during compilation. Computationally expensive loops are accelerated using OpenMP.

Cawkwell, MarcJon↗

DOE OSTI2016

MC3, Version 1

Cawkwell, Marc Jon↗