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DOE OSTI2015

fudge

FUDGEFudge provides (Python based) tools to: • Store and modify nuclear data. • Convert ENDF-6 formatted nuclear data [1] into Generalized Nuclear Data (GND) structure [2]. • Convert LLNL’s ENDL formatted nuclear data [3] into GND. • Plot data from GND files. • Check for physics problems in GND files. • Process nuclear data for use by Monte Carlo and/or deterministic codes. Fudge contains hooks for the future addition of nuclear data processing routines. The Fudge release includes samples of published ENDF-6 formatted data that can be converted to the new GND format. This software is the next release in a series of releases that began with GND version 0.1 [2].

Romano, Paul↗

DOE OSTI2020

Kinetic Simulations of Microstructural Evolution (KSOME)

"Kinetic simulations of microstructural evolution" (KSOME) is an object Kinetic Monte Carlo (OMKC) simulation code. It is developed to simulate microstructural evolution in materials during irradiation. KSOME is developed from ground up with priority given to the flexibility, ease of upgradability, and computation efficiency, in that order. With previously developed simulation codes, various processes (more precisely, the diffusion-reaction processes) that can occur in a system were hardwired into the code. Hence, to perform higher fidelity simulations, a significant coding effort was required every time to incorporate either a more accurate description of existing processes or new processes. With KSOME, the necessity for frequent code upgrades whenever there is a need to perform higher fidelity simulations is either eliminated or reduced significantly. In KSOME, the categories of various diffusion-reaction processes and the data-management system are hardwired, while the execution of the diffusion-reaction processes specific to a system is specified via input files. In the present version of KSOME, the analytical expression required to calculate defect interaction radii are inputted via a text file. However, the ability to parse mathematical expressions from a text file can be extended to input the analytical expression to calculate other defect properties; including the expressions to calculate the change in the migration barriers of mobile defects due to long-range interaction with extended defects

Nandipati, Giridhar [Pacific Northwest National La↗

DOE OSTI2024

SuperNu Version 4.x

We seek to release SuperNu, Version 4.x, as a continuation of development for the open source SuperNu software. The SuperNu, Version 3.x Monte Carlo radiative transfer code for astrophysical transients was released with GPLv3 copyright, asserted by LANL in 2015. For the next release we have features planned for development , including: opacity implementation (including non-local thermodynamic equilibrium effects), generalized source implementation (e.g. for emulating shock heating in Type II supernovae), 3T (electron, ion, radiation) internal energy update, special relativity corrections through O(v^2/c^2), light polarization (e.g. for comparison to spectroplarimetry observations of supernovae and kilonovae), and infrastructure features (checkpoint and restart of simulations, and tools including setup and Slurm batch scripts and simulation post-processing/analysis scripts). These features are intended to improve the fidelity and/or better understand uncertainty in supernova and kilonova light curve calculations.

Wollaeger, Ryan↗

DOE OSTI2025

Microreactor Optimization Using Simulation And Economics (mouse)

Microreactor Optimization Using Simulation and Economics (MOUSE) is a tool that integrates both nuclear microreactor design and reactor economics to provide comprehensive evaluations and optimizations. This tool enables stakeholders to explore the interplay between technical and economic variables, guiding them towards effective and competitive microreactor solutions. For the reactor core simulations, MOUSE leverages the OpenMC Monte Carlo Particle Transport Code to perform detailed core simulations for various microreactor designs. The included OpenMC models are 2D core designs of a Liquid Metal Thermal Microreactor (LMTR), a Gas-Cooled TRISO-Fueled Microreactor (GCMR), and a Heat Pipe Microreactor. Beyond core design, MOUSE includes simplified calculations for: - Calculating the masses of heat exchangers within the system. - Mechanical power of pumps. - Estimating the area occupied by various buildings within the nuclear plant. For the economic analysis, MOUSE provides detailed bottom-up cost estimates, encompassing a wide range of costs including preconstruction costs, direct costs, indirect costs, training costs, financial costs, operation & maintenance (O&M) costs, and fuel costs. These cost estimations are developed using data from the MARVEL project and additional literature sources, enabling the calculation of total capital costs and levelized cost of energy for both first-of-a-kind and nth-of-a-kind microreactors. MOUSE also enables analysis of the cost drivers and competitiveness in the electricity market. MOUSE allows users to modify a wide array of technical and economic parameters to evaluate different scenarios and their impacts. Examples of these parameters include: Fuels, coolants, or reflector materials Enrichment levels Control drum materials and geometry Fuel pin geometry and materials Moderator pin geometry and materials Reactor core and reflector dimensions Packing factor for the TRISO particles Nuclear reactor power and reactor burnup Number of sensors Shielding thickness Reactor vessel and guard vessel dimensions Operational staff requirements Number of emergency shutdowns Levelization period Interest rate Construction duration Since MOUSE is powered by the WATTS toolkit, it supports optimization studies, parametric analyses, and uncertainty calculations/propagation. The optimization techniques enable users to identify optimal design and economic configurations. The parametric analysis tools allow users to explore the sensitivity of various parameters, while uncertainty propagation helps quantify the impact of uncertainties on overall performance and cost. User Interface and Workflow: Currently, MOUSE is a command-line-based tool. Users can input various reactor design or economic parameters, modify the designs, run simulations, and visualize results through comprehensive data visualization and reporting capabilities. The typical workflow involves setting up the reactor model, defining economic parameters, running simulations, and analyzing the results to make informed decisions. By combining advanced design calculations with detailed economic modeling, MOUSE provides a robust framework for optimizing nuclear microreactor technologies, enhancing their competitiveness, and guiding stakeholders towards innovative and cost-effective solutions.

Hanna, Botros [Idaho National Laboratory (INL), Id↗

DOE OSTI2024

Integrating ytopt and libEnsemble to autotune OpenMC

Ytopt is a Python machine-learning-based autotuning software package developed within the ECP PROTEAS-TUNE project. The ytopt software adopts an asynchronous search framework that consists of sampling a small number of input parameter configurations and progressively fitting a surrogate model over the input-output space until exhausting the user-defined maximum number of evaluations or the wall-clock time. libEnsemble is a Python toolkit for coordinating workflows of asynchronous and dynamic ensembles of calculations across massively parallel resources developed within the ECP PETSc/TAO project. libEnsemble helps users take advantage of massively parallel resources to solve design, decision, and inference problems and expands the class of problems that can benefit from increased parallelism. In this paper we present our methodology and framework to integrate ytopt and libEnsemble to take advantage of massively parallel resources to accelerate the autotuning process. Specifically, we focus on using the proposed framework to autotune the ECP ExaSMR application OpenMC, an open source Monte Carlo particle transport code. OpenMC has seven tunable parameters some of which have large ranges such as the number of particles in-flight, which is in the range of 100,000 to 8 million, with its default setting of 1 million. Setting the proper combination of these parameter values to achieve the best performance is extremely time-consuming. Therefore, we apply the proposed framework to autotune the MPI/OpenMP offload version of OpenMC based on a user-defined metric such as the figure of merit (FoM) (particles/s) or energy efficiency energy-delay product (EDP) on Crusher at Oak Ridge Leadership Computing Facility. In conclusion, the experimental results show that we achieve the improvement up to 29.49% in FoM and up to 30.44% in EDP.

Autotuning↗

DOE OSTI2014

Sensitivity Analysis and Data Assimilation in a Subcritical Plutonium Metal Benchmark

Here, this work presents the application of first-order adjoint sensitivity analysis, uncertainty quantification, and data assimilation to a subcritical plutonium benchmark experiment using a modified version of the discrete ordinates radiation transport code Denovo. Previous Monte Carlo simulations of this benchmark saw a consistent overprediction of the mean and variance of the measured neutron multiplicity distribution. It was observed that a small scalar reduction in the value of the 239 Pu-induced fission neutron multiplicity was capable of significantly reducing the discrepancies. This work extends those results by computing first-order sensitivities to each nuclide, reaction type, energy, and material region in the benchmark. The sensitivities are then used in a data assimilationmethodology to simultaneously calibrate all responses and multigroup nuclear data. The resulting best-estimate values for the energy group differential multiplicity $\bar{v}$(E g ) are 1σ to 2σ less than the nominal values found in ENDF/B-VII for energies less than 1.5 MeV.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

An Implicit Correlation Method for Cross-Correlation Sampling, with MCNPX-PoliMi Validation

Monte Carlo particle transport codes used to model detector response are traditionally run in analog mode. However, analog simulations of cross-correlation measurements are extremely time-consuming because the probability of coincident detection is small, approximately equal to the product of the probabilities of a single detection in each detector. The new “implicit correlation” method described here increases the number of correlated event scores, thereby reducing variance and required computation times. The cost of the implicit correlation method is comparable to the cost of simulating single-event detection for the lowest absolute detector efficiency in the problem. The new method is especially useful in the nuclear nonproliferation and safeguards fields for simulating correlation measurements of shielded special nuclear material. The new method was implemented in MCNPX-PoliMi for neutron-neutron cross-correlations with a Cf-252 spontaneous fission source measured by 14 detectors at various angles. The method demonstrated good agreement with analog simulation and reference measurement results. Small differences between nonanalog and analog cross-correlation distributions are attributed to discretization errors that are often not present in practical applications. Improvement in the figure of merit was greater than a factor of 100 in all tested cases.

Marcath, Matthew J.↗

DOE OSTI2022

Studsvik R2 materials test reactor Ad Hoc depletion strategy for the derivation of the fuel isotopic composition of the MPCMIV benchmark

The Ad Hoc Depletion Strategy elaborated by the NINE company, developed in support of the organization of the MPCMIV (Multi-physics Pellet Cladding Mechanical Interaction Validation) benchmark input and output specifications, is presented. This work aims at illustrating the strategy itself and then showing the results obtained with its application over the Studsvik R2 Testing Reactor, which is analyzed in the benchmark. The objective of the application of the strategy is to compute the fuel elements isotopic compositions at the beginning of some core loadings of interest for the benchmark. To this objective, it is necessary to implement first the simulation model of the three single assembly types and perform the infinite lattice depletions, then, to build the full core model and to perform the simulation of the core cycle. All the models and simulations were carried out with the use of the Monte Carlo particle transport code Serpent 2. Finally, the simulations results are assessed against Studsvik isotopic compositions of the fuel elements discharged from the R2 Reactor at the end of the core loading. Several assumptions were necessary during all the steps of the strategy, to overcome the lack of information regarding the core management. For this reason, the solution found at the end of the current analysis may not be completely optimized and further improvements regarding the model assumptions will be tested in a future work. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Brazilian CBP - Technoeconomic analysis data

This data is related to the paper entitled "Techno-economic analysis of sugarcane bagasse and straw conversion into cellulosic ethanol via consolidated bioprocessing". That features the evaluation of sugarcane bagasse and straw conversion to ethanol at stand-alone facilities generating electricity from residues. The following scenarios were evaluated: Conventional, featuring hydrothermal pretreatment, fungal cellulase, and yeast fermentation (current commercial standard); Mid-term consolidated bioprocessing (CBP), relying on bagasse solubilization without pretreatment or cotreatment; and Mature CBP, incorporating cotreatment but no pretreatment and considering significant technological advance of the CBP. Available here are the spreadsheets used for Material and Energy balance calculation, Capital and Operational costs estimation and Cash flow analysis. Also available are the description and python code used for Monte Carlo analysis of the ethanol and capital investment variations. This data can be used as a source to implement other techno-economic analysis in the biorefinary context.

09 BIOMASS FUELS↗

DOE OSTI2011

Graphical User Interface for Simplified Neutron Transport Calculations

A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2012

Description of Axial Detail for ROK Fuel

For the purpose of NDA simulations of the ROK fuel assemblies, we have developed an axial burnup distribution to represent the pins themselves based on gamma scans of rods in the G23 assembly. For the purpose of modeling the G23 assembly (both at ORNL and LANL), the pin-by-pin burnup map as simulated by ROK is being assumed to represent the radial burnup distribution. However, both DA and NDA results indicate that this simulated estimate is not 100% correct. In particular, the burnup obtained from the axial gamma scan of 7 pins does not represent exactly the same 'average' pin burnup as the ROK simulation. Correction for this discrepancy is a goal of the well-characterized assembly task but will take time. For now, I have come up with a correlation for 26 axial points of the burnup as obtained by gamma scans of 7 different rods (C13, G01, G02, J11, K10, L02, and M04, neglecting K02 at this time) to the average burnup given by the simulation for each of the rods individually. The resulting fraction in each axial zone is then averaged for the 7 different rods so that it can represent every fuel pin in the assembly. The burnup in each of the 26 axial zones of rods in all ROK assemblies will then be directly adjusted using this fraction, which is given in Table 1. Note that the gamma scan data given by ROK for assembly G23 included a length of ~3686 mm, so the first 12 mm and the last 14 mm were ignored to give an actual rod length of ~366 cm. To represent assembly F02 in which no pin-by-pin burnup distribution is given by ROK, we must model it using infinitely-reflected geometry but can look at the effects of measuring in different axial zones by using intermediate burnup files (i.e. smaller burnups than 28 GWd/MTU) and determining which axial zone(s) each burnup represents. Details for assembly F02 are then given in Tables 2 and 3, which is given in Table 1 and has 44 total axial zones to represent the top meter in explicit detail in addition to the other 26 zones. Note that the MCNP files for F02 were created using the Monte Carlo burnup linkage code Monteburns, which saves MCNP input files with detailed compositions as a function of burnup. The 'intermediate burnup files' produced for F02 include a cooling time of 27 years. The axial location of 5 spacers was also included in the ROK F02 assembly in which each spacer contained a length of 3.81 cm. Note that due to the nature of Monteburns, which was run in a special fashion for this problem, the step number increments after the 27 year decay, so the second column of Table 2 refers to the step number that should be used in the Monteburns files.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

DHS Summary Report -- Robert Weldon

This summer I worked on benchmarking the Lawrence Livermore National Laboratory fission multiplicity capability used in the Monte Carlo particle transport code MCNPX. This work involved running simulations and then comparing the simulation results with experimental experiments. Outlined in this paper is a brief description of the work completed this summer, skills and knowledge gained, and how the internship has impacted my planning for the future. Neutron multiplicity counting is a neutron detection technique that leverages the multiplicity emissions of neutrons from fission to identify various actinides in a lump of material. The identification of individual actinides in lumps of material crossing our boarders, especially U-235 and Pu-239, is a key component for maintaining the safety of the country from nuclear threats. Several multiplicity emission options from spontaneous and induced fission already existed in MCNPX 2.4.0. These options can be accessed through use of the 6th entry on the PHYS:N card. Lawrence Livermore National Laboratory (LLNL) developed a physics model for the simulation of neutron and gamma ray emission from fission and photofission that was included in MCNPX 2.7.B as an undocumented feature and then was documented in MCNPX 2.7.C. The LLNL multiplicity capability provided a different means for MCNPX to simulate neutron and gamma-ray distributions for neutron induced, spontaneous and photonuclear fission reactions. The original testing on the model for implementation into MCNPX was conducted by Gregg McKinney and John Hendricks. The model is an encapsulation of measured data of neutron multiplicity distributions from Gwin, Spencer, and Ingle, along with the data from Zucker and Holden. One of the founding principles of MCNPX was that it would have several redundant capabilities, providing the means of testing and including various physics packages. Though several multiplicity sampling methodologies already existed within MCNPX, the LLNL fission multiplicity was included to provide a separate capability for computing multiplicity as well as including several new features not already included in MCNPX. These new features include: (1) prompt gamma emission/multiplicity from neutron-induced fission; (2) neutron multiplicity and gamma emission/multiplicity from photofission; and (3) an option to enforce energy correlation for gamma neutron multiplicity emission. These new capabilities allow correlated signal detection for identifying presence of special nuclear material (SNM). Therefore, these new capabilities help meet the missions of the Domestic Nuclear Detection Office (DNDO), which is tasked with developing nuclear detection strategies for identifying potential radiological and nuclear threats, by providing new simulation capability for detection strategies that leverage the new available physics in the LLNL multiplicity capability. Two types of tests were accomplished this summer to test the default LLNL neutron multiplicity capability: neutron-induced fission tests and spontaneous fission tests. Both cases set the 6th entry on the PHYS:N card to 5 (i.e. use LLNL multiplicity). The neutron-induced fission tests utilized a simple 0.001 cm radius sphere where 0.0253 eV neutrons were released at the sphere center. Neutrons were forced to immediately collide in the sphere and release all progeny from the sphere, without further collision, using the LCA card, LCA 7j -2 (therefore density and size of the sphere were irrelevant). Enough particles were run to ensure that the average error of any specific multiplicity did not exceed 0.36%. Neutron-induced fission multiplicities were computed for U-233, U-235, Pu-239, and Pu-241. The spontaneous fission tests also used the same spherical geometry, except: (1) the LCA card was removed; (2) the density of the sphere was set to 0.001 g/cm3; and (3) instead of emitting a thermal neutron, the PAR keyword was set to PAR=SF. The purpose of the small density was to ensure that the spontaneous fission neutrons would not further interact and induce fissions (i.e. the mean free path greatly exceeded the size of the sphere). Enough particles were run to ensure that the average error of any specific spontaneous multiplicity did not exceed 0.23%. Spontaneous fission multiplicities were computed for U-238, Pu-238, Pu-240, Pu-242, Cm-242, and Cm-244. All of the computed results were compared against experimental results compiled by Holden at Brookhaven National Laboratory.

42 ENGINEERING↗

DOE OSTI2014

Monte Carlo Simulations of Neutron Resonance Transmission Analysis with the Dense Plasma Focus

Simulations were performed using the Monte Carlo radiation transport code MCNP in order to analyze the feasibility of using a dense plasma focus to perform neutron resonance transmission analysis. It was found that the dense plasma focus is well-suited for neutron resonance transmission analysis because of the relatively short pulse duration and large neutron yield. This makes lowenergy neutron spectroscopy using time of flight much more accurate, allowing for more detailed assays of smaller quantities of special nuclear material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2014

Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO 2 /zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

Surface Triangulation for CSG in Mercury

Visualization routines for rendering complicated geometries are very useful for engineers and scientists who are trying to build 3D prototypes of their designs. A common way to rapidly add interesting features to a 3D model is through the use of a concept called Constructive Solid Geometry. CSG uses compositions of the boolean set operations to manipulate basic geometric primitives to form more complicated objects. The most common boolean operations employed are union, intersection, and subtraction. Most computer-aided design software packages contain some sort of ability visualize CSG. The typical workflow for the user is as follows: The user specifies the individual primitive components, the user arbitrarily combines each of these primitives with boolean operations, the software generates a CSG tree structure which normally stores these solids implicitly with their defining equation, the tree is traversed and a general algorithm is applied to render the appropriate geometry onto the screen. Algorithms for visualizing CSG have been extensively developed for over a decade. Points sampled from the implicit solids are typically used as input by variations of algorithms like marching cubes and point-cloud surface reconstruction. Here, we explain a surface triangulation method from the graphics community that is being used for surface visualization in the framework of a Monte-Carlo neutron transport code called Mercury.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Incorporation of Spatial Stochastic Variability into Paracousti-UQ

Waves propagating through natural materials such as ocean water encounter spatial variations in material properties that cannot easily be predicted or known in advance. Deterministic wave simulation algorithms must assume that all properties throughout the model space are precisely known. However, a stochastic wave simulation tool can parameterize the material as a stochastic medium with a certain probability distribution and correlation length. This report documents the addition of spatial stochastic variability into Paracousti-UQ, Sandia Geophysics Department's 3-D full waveform acoustic algorithm within stochastic media. The ability of the code to replicate Monte Carlo solutions in 1-D spatially variable media is also evaluated.

58 GEOSCIENCES↗

DOE OSTI2018

Implementation and Validation of Photon Transport in OpenMC

OpenMC is a Monte Carlo particle transport code focused on reactor physics calculations. It stochastically simulates neutrons moving through 3D models using constructive solid geometry and continuous-energy cross sections. Recently, a photon transport capability was added to OpenMC. In reactor simulations, photon transport is needed, for example, to account for photon heating: although the majority of the energy deposited at a fission site comes from the kinetic energy of the fission fragments, a nontrivial portion can be carried away by photons. Therefore, photons produced in neutron reactions must be simulated in order to accurately model energy deposition. OpenMC can now simulate neutron-induced photon production as well as perform pure photon calculations. The four basic photon interactions with matter — coherent (Rayleigh) scattering, incoherent (Compton) scattering, the photoelectric effect, and pair production — are modeled. Secondary processes that can create new photons (i.e., atomic relaxation, electron-positron annihilation, and bremsstrahlung) are also included. The methods implemented here were validated against MCNP6 for a variety of materials and energies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2019

Activation and residual dose rates in the NSLS-II storage ring

It is well known that components in high energy accelerators may get activated and expose workers during maintenance activities. While the potential for activation is higher in positive ion accelerators due to the prolific production of neutrons, certain regions and components in high energy electron accelerators too can be activated enough to create residual dose rates that require careful attention for ALARA compliance. In this report, the activation and residual dose rates from a few materials are studied from various beam loss scenarios ranging from chronic point losses to intermittent real loss scenarios obtained from the operational parameters of the NSLS-II storage ring. FLUKA Monte Carlo radiation transport code is used to model the geometry, the time profile of the beam losses and score the resulting time evolution of the remnant activity and the residual dose rates.

36 MATERIALS SCIENCE↗