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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Extension and Demonstration of NEAMS Multiphysics Tools to Lead-Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 expanded to additional fast reactor applications: lead cooled fast reactors (LFR), sodium cooled fast reactors (SFR), and fast molten salt reactors (MSR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, (2) implementation of a gamma transport capability in the PROTEUS solvers to extend the zooming capability to non-fueled assemblies, (3) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR), and (4) development of a new coupled capability for simulating fast MSRs, specifically modeling precursor transport inside and outside the core. The progress on tasks (1) and (3) are discussed primarily in this document. Tasks (2) and (4) are briefly mentioned in this document, and the interested reader is referred to separate reports for detailed information. First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice, since these tools are targeted to solve many reactor types and geometries. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress and will be completed in FY20. In the fast molten salt reactor (MSR) domain, a new capability was developed to model the flow of neutron precursors both inside and outside of the core, which impacts the effective delayed neutron fraction of the system. The PROTEUS-SN and PROTEUS-NODAL solvers were coupled to Nek5000 in order to simulate precursor flow. The workflows were demonstrated on test problems. The fast MSR modeling capability was supported under two other work packages, and comprehensive details are included in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Analytic evaluation of hydrogen neutron elastic scattering moments with free-gas thermal treatment compared to NJOY

In this article, we present an analytic approach to evaluate Legendre moments of hydrogen neutron elastic scattering differential cross section, with and without free-gas thermal treatment. The results for the zero-th (P 0 ) and first (P 1 ) moments are given and compared against a 138-group cross sections produced by NJOY. An MCNP thin-wire test was employed as an independent verification for our analytic model for the free-gas zero-th moment. Excellent agreement was observed between the analytic result and MCNP prediction. They both suggest that an increased thermal cut-off in NJOY THERMR module is advisable for higher accuracy. Discrepancy was also found between the analytic model and NJOY multi-group cross section for the first order Legendre moments of the differential cross section, which might indicate a numerical anomaly in NJOY.

NJOY↗
DOE OSTI2013

Investigation of the effects of cell model and subcellular location of gold nanoparticles on nuclear dose enhancement factors using Monte Carlo simulation

Purpose: The authors’ aims were to model how various factors influence radiation dose enhancement by gold nanoparticles (AuNPs) and to propose a new modeling approach to the dose enhancement factor (DEF).Methods: The authors used Monte Carlo N-particle (MCNP 5) computer code to simulate photon and electron transport in cells. The authors modeled human breast cancer cells as a single cell, a monolayer, or a cluster of cells. Different numbers of 5, 30, or 50 nm AuNPs were placed in the extracellular space, on the cell surface, in the cytoplasm, or in the nucleus. Photon sources examined in the simulation included nine monoenergetic x-rays (10–100 keV), an x-ray beam (100 kVp), and {sup 125}I and {sup 103}Pd brachytherapy seeds. Both nuclear and cellular dose enhancement factors (NDEFs, CDEFs) were calculated. The ability of these metrics to predict the experimental DEF based on the clonogenic survival of MDA-MB-361 human breast cancer cells exposed to AuNPs and x-rays were compared.Results: NDEFs show a strong dependence on photon energies with peaks at 15, 30/40, and 90 keV. Cell model and subcellular location of AuNPs influence the peak position and value of NDEF. NDEFs decrease in the order of AuNPs in the nucleus, cytoplasm, cell membrane, and extracellular space. NDEFs also decrease in the order of AuNPs in a cell cluster, monolayer, and single cell if the photon energy is larger than 20 keV. NDEFs depend linearly on the number of AuNPs per cell. Similar trends were observed for CDEFs. NDEFs using the monolayer cell model were more predictive than either single cell or cluster cell models of the DEFs experimentally derived from the clonogenic survival of cells cultured as a monolayer. The amount of AuNPs required to double the prescribed dose in terms of mg Au/g tissue decreases as the size of AuNPs increases, especially when AuNPs are in the nucleus and the cytoplasm. For 40 keV x-rays and a cluster of cells, to double the prescribed x-ray dose (NDEF = 2) using 30 nm AuNPs, would require 5.1 ± 0.2, 9 ± 1, 10 ± 1, 10 ± 1 mg Au/g tissue in the nucleus, in the cytoplasm, on the cell surface, or in the extracellular space, respectively. Using 50 nm AuNPs, the required amount decreases to 3.1 ± 0.3, 8 ± 1, 9 ± 1, 9 ± 1 mg Au/g tissue, respectively.Conclusions: NDEF is a new metric that can predict the radiation enhancement of AuNPs for various experimental conditions. Cell model, the subcellular location and size of AuNPs, and the number of AuNPs per cell, as well as the x-ray photon energy all have effects on NDEFs. Larger AuNPs in the nucleus of cluster cells exposed to x-rays of 15 or 40 keV maximize NDEFs.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Automatic MCNP File Generation for Cherenkov Imaging Simulations [Slides]

Overview: Creating physics models to accurately simulate a variety of spent fuel assemblies. These characterized simulations will then be used to create a well documented repository that encompasses a wide variety of different fuel assemblies, burn-up and spent fuel pool conditions, and defects. The final project will use our simulations to train an AI, and for it to be effective it has to have a lot of reliable data; Automating an arbitrary amount of created spent fuel pins with materials, that will be passed into AI models for it to "learn" what a good spent nuclear fuel cell looks like and what a non-spent nuclear fuel cell looks like.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Towards Realistic and High Fidelity Models for Nuclear Reactor Power Synthesis Simulation with Self-Powered Neutron Detectors

As presented in this report, a weighting function–based inferencing method is being applied to synthesize the power distribution in next-generation and university research reactors based on simulated self power neutron detector (SPND) responses. The overall goal is to assess the impacts of sensor uncertainty and true power distribution perturbations on the error in the synthesized power distribution. Regarding sensor uncertainty, the NuScale Small Modular Reactor (SMR) and the Westinghouse AP1000 serve as testbeds for analyzing the impact of varying the sensor uncertainty, as well as varying the number of sensors per sensor string in the reactor core. The reactor models are informed by Monte Carlo N-Particle (MCNP) neutron flux tallies. For the NuScale SMR and Westinghouse AP1000, the SPND response functions (i.e., the response of the SPNDs to individual segments of fuel) were determined homogeneously. Regarding an analysis of power distribution perturbation detection, the Texas A&M Testing, Research, Isotopes, General Atomics Reactor (TAMU TRIGA) reactor was used as a demonstration case with one particular arrangement of SPNDs; the response functions for this reactor model were determined heterogeneously, making this a uniquely high-fidelity demonstration of perturbation detection. Finally, SPND models generated in the Geometry and Tracking 4 (Geant4) code have been generated and tested for comparison with traditionally implemented analytical SPND models, with the intent for Geant4 integration with the full methodological framework. SPND current outputs as a function of distance from some fuel assembly segment in the NuScale SMR are compared with the analytically determined currents. Results from the sensor uncertainty simulations for the NuScale SMR and AP1000 indicate that the average error in the inferred power distribution on the fuel assembly segment level is reasonably low, being slightly less than the random uncertainty applied to all respective SPNDs in both cores. For example, if all SPNDs in the core have a random uncertainty of 5%, then the corresponding fuel assembly segment level error (i.e. difference between the true and inferred local power) is ~2–3%. However, the maximum error in the inferred power distribution on the fuel assembly segment level can be considerably high (>15%) when SPND random uncertainties start to exceed ~3%. In general, the average and maximum errors in the inferred power distribution were slightly higher in the AP1000 as opposed to the NuScale SMR for the sensor string configurations considered herein. Another result determined from analysis of the sensor uncertainty simulations was that increasing the number of SPNDs per string does not clearly reduce inferred power distribution error and can in fact make the error large in some cases; however, this assessment may skewed due to imposed iteration limits. Results from the perturbation detection demonstration using the high-fidelity TAMU TRIGA model indicate that, given the arrangement of 17 SPND strings and 4 SPNDs per string considered herein, there is a clear, provable ability to infer a localized Gaussian-type peak perturbation in the 3D power distribution. Such a perturbation was detected with an average fuel assembly segment level error of 0.19%, and the general visualization of the detected perturbation clearly indicates that the magnitude and shape were appropriately resolved. Finally, the electrical current output generated by the Geant4 modeled SPND indicates significant magnitude differences than the analytically modeled SPND, demonstrating the need for accurate SPND models which account for finite sensor geometry effects to inform the power synthesis work described herein.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Detection of long-lived contaminants in cyclotron-produced radiopharmaceuticals by large area plastic scintillators

In the USA during 2015 over 1,718,500 clinical PET scans were performed in 2380 centers. Patients undergoing these studies will alarm homeland security monitors at borders, ports, and some airports for up to 2 days due to the positron-annihilation photons (0.511 MeV) from the imaging isotope 18 F (110 min half-life). In some radiopharmaceuticals, long-lived contaminants generated from the activation and spallation of elements in the HAVAR window of the cyclotron target may also be present in the patient dose. In this work, we have modeled a typical homeland security portal monitor in MCNP, and determined its sensitivity to each of the possible contaminating isotopes found in PET radiopharmaceuticals when distributed in patient phantoms. Furthermore, these data were combined with the time-dependent uptake/retention functions for the contaminating isotopes, to identify the isotopes that can cause alarms for substantial periods of time after the short-lived imaging agent has decayed away.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2012

Spectra of photons and neutrons generated in a heterogeneous head of a 15 MV LINAC at differents field sizes

Spectra of photons and neutrons were calculated, using the Monte Carlo code MCNP-5 using the e/p/n mode. A heterogeneous model was used to define the linac head where the collimators were modeled to produce five different treatment fields at the isocenter. Photon and neutron spectra were estimated in several points along two directions from the isocenter. The total photon fluence beyond 60 cm behaves according to 1/r{sup 2} rule, while total neutron fluence, beyond 80 cm, can be described by diffusion theory using an infinite plane as a neutron source.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Correction of steel casing effect for density log using numerical and experimental methods in the slim borehole

Density log is widely applied for a variety of fields such as the petroleum exploration, mineral exploration, and geotechnical survey. The logging condition of density log is normally open holes but there are frequently cased boreholes. The primary calibration curve by slim hole logging manufacturer is normally the calibration curves for the variation of borehole diameter. In this study, we have performed the correction of steel casing effects using numerical and experimental methods. We have performed numerical modeling using the Monte Carlo N-Particle (MCNP) code based on Monte Carlo method, and field experimental method from open and cased hole log. In this study, we used the FDGS (Formation Density Gamma Sonde) for slim borehole with a 100 mCi 137Cs source, three inch borehole and steel casing. The casing effect between numerical and experimental method is well matched.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

Validation and Bias Quantification of Criticality Safety Codes for SRNS Operations

The validation and bias quantification methods employed by the Criticality Safety Engineering group at Savannah River Nuclear Solutions are discussed. Both SCALE and MCNP are validated against benchmark experiments, where these benchmark models have been divided into subgroups representative of the types of fissile material operations at SRNS. A quantification of the bias is made for each subgroup for each code and the three methods employed for quantification are discussed. The bias information is employed directly in the establishment of the Upper Safety Limit (USL) value of the neutron multiplication factor when using standardized documented SRNS methods. A simple example of validation and bias quantification is given and the procedure for setting the USL using the derived bias information is also presented.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Utilizing Sensitivity and Correlation Coefficients from MCNP and Whisper to Guide Microreactor Experiment Design

When designing experiments for full-scale reactor systems, MCNP®* and Whisper can be used to create neutronic models and compare the similarity of two nuclear systems via correlation coefficients for κ eff , effective multiplication factor. This thesis applies this framework to a conceptual heat-pipe, yttrium-hydride moderated microreactor system and experiments. The framework is intended as a supplement to other neutronics/thermal/multiphysics analyses and provides a concrete method to measure the neutronic similarity of two systems. By analyzing the shared nuclear data uncertainty, as well as sensitivity to nuclear data over all neutron energies, highly informative experiments can be designed to aid in the development of microreactor and other advanced reactor technologies and systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

AR reactor neutronics multi-code validation*

The affordable, robust, compact (ARC) reactor is a tokamak fusion reactor concept currently under development by Commonwealth Fusion Systems and Massachusetts Institute of Technology. There are three important neutronics considerations for the operation of the ARC reactor: (1) breeding of enough tritium in the blanket to sustain the D–T reaction in the plasma; (2) ensuring low fluence on the superconducting toroidal field coils; and (3) assessing neutron volumetric heating in structural components. Here, this work aims to perform a validation of the neutronics analysis approach by code-to-code comparison. State-of-the-art software stacks are employed for the neutronics analysis of the ARC reactor, and a computer-aided design (CAD) model is used directly for Monte Carlo (MC) neutron transport calculations. Three software stacks, Attila-MCNP, OpenMC-DAGMC, and Shift-DAGMC, are used to perform neutronic analyses of a 90° sector CAD model of the ARC reactor. In conclusion, results show that the flux tallies calculated by the three software stacks are very close. Also, the volumetric heating and tritium breeding values have less than 0.6% relative difference between codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Algorithmic Improvements to MCNP5 for High-Resolution Fusion Neutronics Analyses

Neutronics analyses of the ITER experimental fusion reactor rely on increasingly complex geometry models and estimates of energy-dependent neutron flux and radiation dose-rate distributions generated at ever higher resolutions. There are significant practical challenges with applying the Monte Carlo N-Particle (MCNP) continuous-energy transport code to high-resolution analyses. For models consisting of more than 100 000 surfaces and cells, geometry initialization can take several hours, thus slowing down model integration and transport analysis efforts. In multithreaded simulations, the amount of memory consumed by superimposed mesh tally data increases in proportion to the number of threads. This behavior limits either the tally resolution or the number of processor cores that can be utilized in the simulation. This paper describes algorithmic improvements that were implemented in a modified version of MCNP5 to overcome these limitations. These improvements are referred to as the Oak Ridge National Laboratory Transformative Neutronics (ORNL-TN) upgrade. Here, a comparison of the performance and memory usage of both MCNP5 and ORNL-TN on several relevant fusion neutronics models is presented. In these tests and in actual high-resolution neutronics analyses, ORNL-TN reduces geometry processing times from hours to a few seconds and increases in-memory mesh tally capacity from the order of 10 8 to 10 10 space-energy bins.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Opportunities for Undergraduate Research in Nuclear Physics

University of Dallas (UD) physics majors are offered a variety of undergraduate research opportunities in nuclear physics through an established program at the University of Kentucky Accelerator Laboratory (UKAL). The 7-MV Model CN Van de Graaff accelerator and the neutron production and detection facilities located there are used by UD students to investigate how neutrons scatter from materials that are important in nuclear energy production and for our basic understanding of how neutrons interact with matter. Recent student projects include modeling of the laboratory using the neutron transport code MCNP to investigate the effectiveness of laboratory shielding, testing the long-term gain stability of C 6 D 6 liquid scintillation detectors, and deducing neutron elastic and inelastic scattering cross sections for 12 C. Finally, results of these student projects are presented that indicate the pit below the scattering area reduces background by as much as 30%; the detectors show no significant gain instabilities; and new insights into existing 12 C neutron inelastic scattering cross-section discrepancies near a neutron energy of 6.0 MeV are obtained.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

CADIS and FW-CADIS Variance Reduction in Gamma Transport for Predicting Prompt Forensics Signatures

The goal of prompt nuclear forensics is to determine the characteristics of a nuclear detonation based on the signatures available almost immediately after the explosion. An important characteristic is the reaction time history (RTH), a measure of the device’s rate of neutron multiplication. The RTH can be estimated by observation of the gamma radiation emitted from the detonation, which can be detected directly or observed indirectly as Teller light. Gamma transport simulations used to predict these radiation fields are often modeled stochastically using the Monte Carlo N-Particle (MCNP) code, which can be a computationally demanding task due to the number of particle histories needed to achieve statistical convergence. In an attempt to improve the efficiency of these calculations, we evaluate two variance reduction techniques: Consistent Adjoint-Driven Importance Sampling (CADIS) and Forward-Weighted Consistent Adjoint-Driven Importance Sampling (FW-CADIS). These methods use a deterministically calculated adjoint flux to create weight windows and source biasing that guide MCNP sampling. We study the utility of CADIS and FW-CADIS for their use in MCNP gamma transport for nuclear forensics prediction simulations. Furthermore, the results demonstrate that both CADIS and FW-CADIS improve the accuracy for forensics-focused simulations, with CADIS being most beneficial in direct detection and FW-CADIS being ideal for computing a global Teller light source.

CADIS↗
DOE OSTI2018

Neutronics Analyses for the Spallation Neutron Source Second Target Station

The Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR), two of the word-class neutron scattering facilities, are located at the Oak Ridge National Laboratory. The SNS and HFIR are funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Science, and are operated as user facilities, available to researchers from all over the world. Currently there are thirteen neutron scattering instruments in operation at the HFIR and twenty at the SNS First Target Station.Producing the cutting edge science requires continuous improvements and development of the facilities and instruments. The SNS was designed from the beginning to allow addition of the Second Target Station (STS), and an upgrade of the accelerator power. At this time both advances: the accelerator upgrade and the construction of the STS are in preparation.Current baseline design for the STS calls for a rotating compact tungsten target. The target station is driven with short (<1 μs) proton pulses at 10 Hz repetition rate and 467 kW proton beam power, and is optimized for high intensity and high resolution long wavelength neutron applications. The proton beam footprint as small as acceptable from the mechanical and heat removal aspects is planned to generate a compact-volume neutron production zone in the target, which is essential for tight coupling of the target and the moderators and for achieving high-intensity peak thermal and cold neutron fluxes. The STS will allow operation of approximately 22 beamlines and will expand and complement the current national neutron scattering capabilities.This paper will present an update of the status of the neutronics analyses performed for the STS and will discuss the performance of the moderators, heating rates, activation analyses, and shielding calculations which provide input in the engineering design.

Remec, Igor↗
DOE OSTI2016

SU-F-T-183: Design of a Beam Shaping Assembly of a Compact DD-Based Boron Neutron Capture Therapy System

Purpose: To design a beam shaping assembly (BSA) to shape the 2.45-MeV neutrons produced by a deuterium-deuterium (DD) neutron generator and to optimize the beam output for boron neutron capture therapy of brain tumors Methods: MCNP is used for this simulation study. The simulation model consists of a neutron surface source that resembles an actual DD source and is surrounded by a BSA. The neutron source emits 2.45-MeV neutrons isotropically. The BSA is composed of a moderator, reflector, collimator and filter. Various types of materials and geometries are tested for each component to optimize the neutron output. Neutron characteristics are measured with an 2×2×2-cm{sup 3} air-equivalent cylinder at the beam exit. The ideal BSA is determined by evaluating the in-air parameters, which include epithermal neutron per source neutron, fast neutron dose per epithermal neutron, and photon dose per epithermal neutron. The parameter values are compared to those recommended by the IAEA. Results: The ideal materials for reflector and thermal neutron filter were lead and cadmium, respectively. The thickness for reflector was 43 cm and for filter was 0.5 mm. At present, the best-performing moderator has 25 cm of AlF{sub 3} and 5 cm of MgF{sub 2}. This layout creates a neutron spectrum that has a peak at approximately 10 keV and produces 1.35E-4 epithermal neutrons per source neutron per cm{sup 2}. Additional neutron characteristics, fast neutrons per epithermal neutron and photon per epithermal neutron, are still under investigation. Conclusion: Working is ongoing to optimize the final layout of the BSA. The neutron spectrum at the beam exit window of the final configuration will have the maximum number of epithermal neutrons and limited photon and fast neutron contaminations within the recommended values by IAEA. Future studies will also include phantom experiments to validate the simulation results.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Validation and bias quantification of criticality safety codes for SRNS operations

The validation and bias quantification methods employed by the Criticality Safety Engineering group at Savannah River Nuclear Solutions (SRNS) are discussed. Both SCALE and MCNP are validated against benchmark experiments, where these benchmark models have been divided into subgroups representative of the types of fissile material operations at SRNS. A quantification of the bias is made for each subgroup for each code and the three methods employed for quantification are discussed. The bias information is employed directly in the establishment of the Upper Safety Limit (USL) value of the neutron multiplication factor when using standardized documented SRNS methods. A simple example of validation and bias quantification is given and the procedure for setting the USL using the derived bias information is also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

MCNP6 and DRiFT modeling efforts for the NEUANCE/DANCE detector array

This report seeks to study and benchmark code predictions against experimental data; determine parameters to match MCNP-simulated detector response functions to experimental stilbene measurements; add stilbene processing capabilities to DRiFT; and improve NEUANCE detector array modeling and analysis using new MCNP6 and DRiFT features.

47 OTHER INSTRUMENTATION↗