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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

MCNP6 Study of Spatial Fidelity Required for ROC Curve Convergence

The detection of special nuclear material at ports and borders is one of the primary missions of the Department of Homeland Security's (DHS) Domestic Nuclear Detection Office (DNDO). In support of this mission a large-scale modeling and simulation effort has been undertaken with the goal of providing a comprehensive tool-kit that addresses the complexity of problems involved with developing and maintaining technologies deployed to global nuclear detection architecture (GNDA) nodes. An important objective of DNDO's mission is to characterize detector system performance before deployment. In the effort to support this mission, the latest version of MCNP has been used to perform detectability studies of highly enriched uranium (HEU) to populate a database that will be used for the tool-kit. The characterization method used in the implementation of the database is the receiver operating characteristic (ROC). A ROC curve shows the limits of detectability of a given system as well as an overall metric of the efficacy of a detection system. This is accomplished by comparing total detector count distributions to the count distribution produced by background alone. For an in-depth explanation of the theory behind the MCNP production of ROC curves please refer to the paper, 'How To Use The Receiver Operating Characteristic Tally Option in MCNP6'. Building a useful database requires an expansive parameter space. The current tool-kit use nine parameters, which if varied with just 5 options each, results in nearly two million simulations to run. Depending on the set of parameters these simulations can take a few hours or a few weeks to complete. This computational cost necessitates the use of scaling algorithms to interpolate between the data contained in the database. The tool-kit employs a discrete event method to simulate the movement and detector response of special nuclear material (SNM) through a GNDA node. Given the nature of a discrete event simulation, constant motion of SNM must be split up in to discrete spatial detection points. The fidelity required of the spatial splitting to achieve a properly converged answer is investigated here for a neutron detection system. The detector used was a SNAP-3 neutron detector in front of a sphere of SNM in the presence of background radiation. A parameter study has been performed to determine the spatial fidelity required for ROC curve convergence for moving objects in a discrete event simulator. It appears that 5 points is the minimum number of detection points with 7 points needed for one of the parameter sets. The convergence worsens with increases to detection length but appears to improve with increasing detector distance. These results will help future work involving quantifying the accuracy based on given tolerances and parameters. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Predicting Future Solar Modulation and Implementation in MCNP6

The application of detecting low level radiation in the field requires quantification of the natural radiation background specific to that location. This background is made up two primary components: naturally decaying radionuclides in the soil and atmosphere known as terrestrial background, and the particle cascades caused by high energy cosmic radiation interacting within the earth's atmosphere known as cosmic background. The terrestrial component dominates the background spectrum below approximately 2 MeV, however, the cosmic component comprises the entirety of the high-energy spectrum. Furthermore, the terrestrial component is relatively constant for a given location while the cosmic component is heavily dependent on solar activity, an effect known as solar modulation. Annual values of the solar modulation parameter, φ, have been determined since 1936, and their fluctuations are seen to be roughly sinusoidal over time. If one wished to predict or simulate the outcome of an experiment in the future, a prediction for the solar modulation would be necessary to accurately determine the cosmic background contribution. The previous version 6.1.1 of MCNP contained solar modulation data for 46 years between 1960 and 2005. This data was updated with newer solar modulation data, from I. G. Usoskin et al., which reports solar modulation data for an additional 33 years between 1936 and 2014. In order to use past solar modulation data to predict cosmic background radiation fluctuations for dates beyond 2014 an algorithm was developed to forecast future solar modulation data based on the sinusoidal behavior exhibited in recorded data. This algorithm was implemented into MCNP 6.1.2 to be used for cosmic background simulations and tested against benchmark data to assess its accuracy. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Characterizing the Thermal Column Beam in the University of Massachusetts Lowell Research Reactor

The University of Massachusetts Lowell operates a 1 MW, light water cooled and moderated Research Reactor (UMLRR) fueled by low-enriched uranium to generate neutron flux that is used at a variety of in-core and beam port facilities within the reactor to perform irradiation experiments for academic and industrial research. One of the irradiation sites is a graphite-moderated thermal column designed to deliver a collimated beam of thermal neutrons to a neutron radiography imaging system. A detailed characterization of the neutron beam is essential for the design and interpretation of experiments at the facility and for radiation safety. This work aimed to characterize the neutron flux and energy spectrum in the UMLRR thermal column beam by employing experimental and computational methods known as the foil activation and spectrum unfolding technique. A previously existing Monte Carlo N-Particle transport (MCNP) model of the UMLRR reactor core has been extended to the thermal column where the neutron flux and energy spectrum have been computed and experimentally validated. This work employed the foil activation and spectrum unfolding technique along with neutron transport computation to extend the knowledge of the neutron fluence and energy spectrum from the UMLRR core to the adjacent thermal column beam. Radiography systems are used for materials characterization, studying dynamic behavior of engineering systems, and for nuclear forensics applications. Foil activation is an experimental method where thin foils of highly pure materials are placed in the beam where interactions between the neutrons and the foil material activate radioisotopes in the foils whose activities can be measured after the irradiation. Downstream computation with MCNP and the spectrum unfolding code SAND can deconvolute the neutron fluence and energy spectrum that must have been necessary to induce the production rates of radioisotopes that were obtained through foil activation. The necessary ingredients for successful spectrum unfolding are the experimentally measured production rates and some a priori (initial guess) knowledge of the flux spectrum. The thermal column is equipped with a neutron radiography imaging system, so any experimenters who desire to use neutron radiography at UMass Lowell will benefit significantly from knowing the characteristics of the neutron beam. Comprehensive knowledge of these characteristics of the beam is essential to compare and benchmark results obtained for similar experiments at other facilities. The Integrated Nuclear Security and Safeguards Laboratory (INSSL) is pursuing the investigation of the performance of self-healing materials under stress-strain environments and the characterization of radiation detectors and special containers for storing and transporting nuclear material. This work contributes to the improvement of the current radiography system, expands the utility of the system for nuclear safeguards and security application and provides information that will enhance the general safety of the UMLRR experimental facilities. The main objective of this work has been achieved; an appropriate neutron flux spectrum for the UMLRR thermal column beam has been determined using the foil activation method followed by spectrum unfolding. There now exists an experimentally validated computational model of the neutron flux spectrum and the thermal column's neutron environment is characterized. We can conclude that the thermal column is well characterized; an appropriate neutron flux spectrum has been determined through an experimentally validated computational model. Future experimenters interested in the thermal column may now expand on the computational model or use the spectral information for their own purposes. The knowledge gained herein will be helpful for the INSSL's future work using neutron radiography to analyze the effectiveness of self-healing materials and special containers. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Acceleration Techniques using ADVANTG for the MDF Design and Optimization

The Spallation Neutron Source (SNS) facility at Oak Ridge National Laboratory (ORNL) is implementing a Moderator Demonstration Facility (MDF) to demonstrate the performance characteristics of advanced moderators central to the Second Target Station (STS) for SNS. The MDF will provide a small, short-pulsed neutron source for slow neutron production in which we intend to easily and quickly change from one moderator concept to another, characterizing the performance of each moderator in a reproducible and consistent fashion. The MDF will be unique and innovative amongst the world's moderator test facilities in that it will be intended to be very prototypic in terms of 'moderator illumination' - the spatial variation of the neutron flux entering the moderator itself - as well as capable of testing so-called high-brightness moderators in a wing configuration. During the design phase of the MDF, the Monte Carlo N-Particle transport code MCNPX v.2.7.0 is being used as the primary tool for the neutronic analysis of the facility and characterization of the moderator performance. The Automated Variance Reduction Generator (ADVANTG) code is being utilized to accelerate the MCNPX calculations. The ADVANTG code generates space- and energy-dependent weight window parameters for a successive MCNP run by calculating the adjoint flux with the Denovo discrete ordinates code based on the user defined MCNP tally. We have tested various strategies using ADVANTG to generate weight window parameters to efficiently calculate detailed moderator performance. Using point detector response functions for weight window parameter optimization with ADVANTG did not work well for improving the efficiency of Monte Carlo calculations of neutron emission into beam lines. However, using cell fluxes of the area of highest contributions to the point detector for the variance reduction parameter analysis worked very well and improved the FOM of the calculation of neutron emission into a beamline by a factor of 50 over the unbiased case. (authors)

42 ENGINEERING↗
DOE OSTI2020

ISOCS-based Extended In Situ Gamma Spectrometry Services Tool SuperISOCS for Waste Measurements - 20275

Canberra's In Situ Object Counting System (ISOCS) is an established tool used for gamma spectrometry. It allows physical representation of complex geometries and mathematical calculation of the calibration function while avoiding the need for radioisotope standards. ISOCS provides a large range of templates to generate the geometry for most nuclear waste components in nuclear facilities. Nonetheless, there are some complex geometries where the accuracy attainable is limited by the conventional ISOCS templates. An ISOCS-based extended in-situ Gamma Spectrometry service tool has been developed, based on the so-called SuperISOCS (SISOCS) software. This tool was developed to generate the most complex geometries. The principle of geometry construction is based on the composition of 8 different primitive geometries. The primitive object is able to be cut and to change a part of an object. An object is able to overlap to another one to form a more complex object. These functions refine the modelled geometry compared to the real objects and decrease the measurement systematic error because they allow the mathematical model to represent the real-life geometry more faithfully through reducing the approximations and simplifications that need to be made The algorithm of SISOCS is the same as that of standard ISOCS. A set of calculations was done in comparison with MCNP efficiency calculation and SISOCS results. 80 different models were compared between SISOCS and MCNP, with added 30 cases compared between ISOCS and SISOCS. The average differences were less than 2%, for all energy's efficiency calculations. In this paper we illustrate the benefits that this new tool can provide, in terms of improved accuracy for complex realistic geometries, through simulation of some example scenarios. Furthermore, we demonstrate, through SISOCS modelling of the real object data test, the comparison between SISOCS to ISOCS calculations for the same geometry. This paper describes the added functions that are easily employed using SISOCS. Our results show the potential application and benefit of this technique for real waste assay projects and the way to integrate the tool into the waste assay systems. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Multi-Physics Coupling to Model a RbCl-RbCl-Ga Target Stack

A computational model has been developed to simulate the performance of the Los Alamos Isotope Production Facility's (IPF) routine production target configuration, which consists of two Inconel-encapsulated RbCl salt targets and one niobium-encapsulated gallium target separated by water cooling channels. The methodology involves iteratively coupling particle transport and thermal hydraulic codes to determine the time dependent behavior of the target. Phase and density changes seen in targets are known to impact beam penetration and distribution. In addition, evidence of energy distributions seen in experimental radionuclide yield measurements at IPF do not match those seen in previous MCNP predictions. Density reduction as a result of heat deposition affects proton energy loss along the particle's track, impacting both radionuclide yields and target design. Accurate modeling of the thermal behavior of these multi-phase targets and the corresponding impact on beam behavior will significantly improve our understanding of incident and exit proton energy distributions and resulting radionuclide yields in downstream targets. Thermal hydraulic modeling, including computational fluid dynamics (CFD) and heat transfer, was performed with ANSYS CFX while particle transport was performed with the Monte Carlo N-Particle code (MCNP). (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Thermal Hydraulics Safety Analyses of TRIGA Reactor Upon Fuel Reconfiguration

Safety analyses of the Training, Research, Isotope, General Atomics (TRIGA) reactor at Kansas State University(KSU) is performed using thermal hydraulic simulations based on changes to reactivity or the volumetric heat generation rate due to changes in the positioning of the enriched fuel elements in the reactor core. If the reactivity or neutronics parameters are known, then they can be used to simulate the thermal energy state of the material which is crucial to assess the safety of the nuclear fuel inside a nuclear reactor. Several thermal hydraulics models and codes have been developed in the past to assess reactor safety. RELAP (Reactor Excursion and Leak Analysis Program) is one such tool widely used throughout the world for safety analyses. Although RELAP is considered a reliable tool by regulators, vendors, and operators, it lacks the numerical fidelity provided by 3-D computational fluid dynamics (CFD) models. This work involves the critical task of assessing safety of a research reactor upon fuel reconfiguration based on models developed in RELAP and comparing them with a commercially available 3-D CFD program. The TRIGA Mark-II reactor, designed by General Atomics and located at Kansas State University, is a class of small pool type inherently safe research reactor licensed to operate at steady state power of up to 1.25 MW. It has a cylindrical core measuring 0.45 meters diametrically and 0.38 meters axially positioned in a tank measuring 1.98 meters diametrically and 6.25 meters axially. It has uranium zirconium hydride (UZrH) fuel rods housed in stainless steel cladding material arranged in five concentric rings (B,C,D,E and F rings, in that order, moving outwards from the center). TRIGA fuel is nominally 20% enriched, but the reactivity of the fuel can be changed by using burnable poisons (not used at KSU), or by varying the total weight percent of uranium in the UZrH fuel meat. The TRIGA reactor is designed to be an inherently safe reactor and is designed to be cooled by natural convection. Thus it is of paramount importance to avoid the film boiling regime; failing to do so will subject the fuel to severe conditions, possibly resulting in its failure due to fuel/clad interactions. A high DNBR ratio is a primary consideration in the reactor core design and thus any modification in the core should ensure that this criteria is always met. In order to fully utilize the capabilities of KSU TRIGA reactor facility, it should be operated at its full capacity, i.e. 1.25 MW, which requires increasing the fuel loading in some of the E- and F-ring elements from 8.5% U (nominal) to 12.0% U (nominal). Only four elements are to be switched to 12% loading, causing local power peaking in these elements. The degree of power peaking to be expected in these 12%-loaded fuel elements was calculated based on a detailed MCNP (Monte Carlo N-Particle code) model of the reactor core. Some preliminary results of the calculations are presented in this paper based on changes in peaking factors and power levels. To predict reactor thermal behavior, axial changes in the peaking factors obtained from MCNP models were used to calculate volumetric energy density functions for a single channel one dimensional RELAP model and a three dimensional COMSOL model of the TRIGA MARK-II reactor. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Characterization of the Pulsed Radiation Environment in FREC-II at the ACRR

Ongoing tests at the Annular Core Research Reactor (ACRR) at Sandia National Laboratories aim to measure temperature and strain in materials exposed to a transient neutron-photon environment. Radiation transport and finite element analysis codes are used to predict the thermal-structural response of test articles exposed in the reactor. Strain and temperature sensors fielded on test articles provide validation data for finite element models, while active and passive radiation dosimetry provides data for MCNP models. The ACRR is able to provide pulse energies of 40 - 300 MJ per pulse with pulse times of 7 - 100 milliseconds. The pulse power is higher than many of the other research reactors due to its unique high heat capacity UO{sub 2}-BeO fuel in the central reactor. Additionally, the Fuel Ringed External Cavity (FREC-II) contains U-ZrH fuel. Future tests at the ACRR plan to field test assemblies in the Fuel Ringed External Cavity (FREC-II) and expose the test articles to the transient anisotropic radiation field experienced in the cavity. The FREC-II is coupled with the ACRR core and is used to test larger specimens, and the radiation environment can be shaped/modified to the needs of the experiment. The FREC-II was designed to provide an irradiation cavity with a strong flux gradient favoring the central cavity side. Neutron fluence also varies by height in the cavity, with a peak level slightly below the fuel rod vertical centerline. Tests were conducted in 2014 to measure the thermal-structural response of test articles exposed in the central cavity. Those tests in the central cavity fielded Nickel activation foils to record the neutron fluence at various locations around the test assembly. Foil data was later used to highlight uncertainties between thermo-structural response models and the measured temperature and strain. Studies performed in 2015 exposed large arrays of activation foils and thermo-luminescent dosimeters (TLD) in the FREC-II cavity in order to map out the neutron and photon spectrum at many locations. This type of passive measurement was shown to be too simple for performing thermo-structural simulations of a test article in response to multiple pulsed-reactor exposures. The Depleted Uranium Strain and Temperature - Enhanced Response (DUSTER-II) experiments will expose test assemblies to reactor pulses in the FREC-II cavity. To prepare for these experiments, the MCNP model will be compared with June 2016 experimental results from the pulses with the FREC-II coupled with the ACRR. These comparisons will provide an accurate model to be implemented in further studies. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Investigations into alternative radiation transport codes for ITER neutronics analysis

As radiation transport models for ITER become increasingly complex, the use of traditional constructive solid geometry (CSG) modelling approaches such as MCWNP presents significant disadvantages during model creation and integration. Studies were performed to assess the capabilities of alternative mesh-based neutronics workflows offered by the MCNP, DAG-MCNP and Serpent-2 codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Calculation of Activated Water Radiation Maps in ITER

ITER is a nuclear machine designed for 500 MW DT fusion power emitting an immense power in the form of 14 MeV neutrons (∼400 MW). These high energy neutrons activate the water used to cool the first wall, blanket modules, vacuum vessel etc. The principal isotopes generated in water coolant are {sup 16}N and {sup 17}'N from oxygen isotopes. Both these radioisotopes are generated by neutrons having energy more than 10 MeV. Even though these isotopes are short-lived (T{sub 1/2} {sup 16}N: 7.13 s and T{sub 1/2} {sup 17}N: 4.17 s) they generate significant radiation fields outside the bio-shield due to high neutron fluence at irradiation end, large volume of the coolant, high speed of the coolant, emission of high energy nuclear radiations from {sup 16}N (gammas) and {sup 17}N (neutrons). Therefore, radiation fields generated by the activated water need to be calculated in order to design the plant systems to withstand radiation dosage and for the safety of the workers as per regulatory limits. This paper presents the calculation methodology of the {sup 16}N and {sup 17}N sources in the coolant lines, manifolds, heat exchangers etc., which are distributed over the wide area outside the bio-shield. The calculated source is modelled in the MCNP building model which is approximately 120 m x 80 m x 50 m. The building model has significant amount of shielding walls (∼3 m thickness). This presents a challenge to get the sufficient statistics everywhere in the global model. The radiation transport is performed with the code MCNP to calculate the radiation fields (biological dose rate, absorbed dose in silicon, 1 MeV silicon equivalent fluence, neutron flux) throughout the tokamak complex.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Study of Whisper Statistics for Criticality Safety Code Validation

Whisper is computational software designed to assist with validation studies with the Monte Carlo particle transport package MCNP. Historical approaches rely on expert based judgment to select benchmarks while Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application model. Whisper will also derive a calculational margin and margin of subcriticality (MOS) that may be used to generate an upper subcritical limit (USL) for a particular application model. Whisper finds benchmark experiments whose isotope sensitivity profiles align best with the MCNP model in question, and assign a correlation coefficient ('ck' in Whisper parlance) to each selected benchmark as a measure of how closely those profiles agree. When ck values are high, there is less uncertainty between model and benchmarks and a higher USL is possible, relative to cases when ck values are low. The default Whisper settings recommends a 100,000 particles per cycle run for 500 active cycles, resulting in 50 M active neutron histories per application run. This is computationally intensive and may not be necessary to produce accurate results. A Whisper statistical study was done varying the neutron histories per cycle from 10,000 to the default 100,000. A lower number of neutron histories will allow Whisper to be used on low memory systems such as laptops. The intent of this paper is to present the changes in the Whisper results as the number of neutrons histories run is varied.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Absolute Dosimetry Model of the University of Washington Clinical Neutron Therapy System (CNTS)

The University of Washington (UW) Clinical Neutron Therapy System (CNTS) generates a fast neutron beam using 50.5 MeV protons incident on a beryllium target. The CNTS is mainly used for fast neutron therapy in the treatment of cancer but is also used for radiation effects testing and the production of rare isotopes, or isotopes with a short half-life, for medical imaging and therapy. The accelerator and treatment head is modeled using the MCNP6 Monte Carlo code to enable beam characterization (neutron and photon fluence, linear energy transfer, secondary particle production), study the behavior of neutron dosimeters and ionization chamber (IC) responses, and to conduct first principle studies of the relative biological effectiveness (RBE) of fast neutrons compared to {sup 60}Co γ-rays. Measurements of neutron dose in selected CNTS fields were made using a model-17 ion chamber (IC-17) flowing methane-based, tissue-equivalent (TE) gas. As a first approximation, we compared IC-17 measurements to the MCNP calculated dose in a thin layer of water at selected locations along the central axis of the beam. To examine the impact of the composition of the chamber fill-gas and the chamber design characteristics, we also developed a high-resolution (∼ 50 μm) MCNP model of the IC-17 from a computerized tomography (CT) scan. Initial results indicate that the dose response of the IC-17 ion chamber is up to 30% larger (smaller?) compared to the dose to water for neutrons with energies >20 MeV. Differences in chamber and water response as large as 800% occur for neutrons with kinetic energies below 1 keV. The simulations of IC-17 response (and dose to water) for CNTS fast neutron spectra match measured doses within 4.6% for all field sizes and irradiation geometries considered. This reported studies provide additional evidence establishing use of the IC-17 ion chamber model or a thin layer of water as a reasonable surrogate for CNTS fast neutron dose calculations.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Validation of nuclear data based on the ASPIS experiments from the SINBAD database - 25428

SINBAD, the Shielding Integral Benchmark Archive and Database, includes the experimental data relative to integral benchmark experiments relevant for radiation shielding applications. The ASPIS experiments, included in the Reactor Shielding section of SINBAD, provide neutron activation foils and spectra measurements performed before/between shielding slabs of Graphite, Sodium, Iron, Steels, Water and combinations of them. Benchmark models for the measurement simulations with the MCNP code are usually provided in SINBAD. Nonetheless, a user-friendly tool is presented, which would be useful for an independent production of MCNP benchmark models. It is based on GPL C++ code called CombLayer. The reports issued for SINBAD on the quality assessment of the ASPIS experiments are also described. The discussion on the quality of the ASPIS experiments for their use in nuclear data evaluations had a positive outcome, except for two experiments out of ten. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Update of the nuclear criticality slide rule calculations: studies with common shielding materials - 25467

This paper presents preliminary results from the third phase of the current update to the nuclear criticality slide rule, in which MCNP, MAVRIC/Monaco, and COG have simulated the prompt neutron and photon dose from critical spheres of 4.95% enriched uranyl fluoride solution and 93.2% enriched uranium metal. New to this phase is the inclusion of various thicknesses of lead, steel, concrete, and water shielding. This phase also evaluated the effects of humidity on the unshielded configurations in the first phase, as well as the effects of changing the ground composition used in the first phase from concrete to dry soil. The small sample of results presented show that MCNP, MAVRIC/Monaco, and COG results are statistically the same or different by only a few percent. The impact of the shielding materials depends on their thickness, but doses decreased by nearly a factor of 100 are presented. In a few instances, buildup of gamma doses occurs due to neutron interactions in shielding materials. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

An analysis of various solution strategies and perturbations on inputs of the reactor shielding problem - 25584

Previous work focused on the use of automated variance reduction parameters for use in a fixed source reactor shielding problem to solve for ex-core particle transport results. The current work assesses that strategy in light of a fully-coupled Monte Carlo problem run as a critical system with parallel processing and the use of moderate variance reduction parameters set manually. The results from the two strategies compare well. Also, results are compared between the MCNP, Virtual Environment for Reactor Applications (VERA) and Shift computer codes. Shift and VERA deliver similar results and compare well to MCNP. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Testing the Delayed Gamma Capability in MCNP6 - Paper 86

The mission of the Domestic Nuclear Detection Office (DNDO) is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material (SNM) for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism. MCNP6 is capable of transporting the various types of radiation encountered in these scenarios; however, the accuracy of MCNP6 must be quantified such that we can be confident in the results of the simulations. Delayed gamma transport is one important component of these detection scenarios. Delayed gamma transport is defined as the combination of gamma emission from radioactive and subsequent transport through matter. MCNP always possessed the ability to emit delayed neutrons from fission; however, these delayed neutron spectra and intensity were functional fits and not directly sampled from the created fission products. Furthermore, there existed no method to emit the decay photons from radioactive fission products (or any radioactive nucleus) A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. At the point of capture, a residual nucleus is created and decayed using the CINDER90 depletion code. MCNP only sends CINDER90 a residual radioactive nucleus and not a flux (i.e. capture rate). As a result, the Bateman equations for the solution are coupled first order differential equations with constant coefficients. Theoretically, these equations could be solved analytically; however, CINDER90 employs a numerical Markov linear chain scheme to solve these equations. This paper addresses a more simplified approach to benchmarking in order to best assess: (1) whether or not deficiencies exist in the current methodology for certain types of decay chains; and (2) paths forward for improving the delayed particle emission capability. We utilize closed form analytic solutions to the granddaughter equations for particular sets of isotope systems to compare the accuracy of the delayed gamma capability in MCNP6. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Rigorous Two-Step Activation for Fusion Systems with PyNE - Paper 127

In fusion systems, high energy neutrons activate system components. After shutdown, the biological dose rate from photons emitted from these components must be estimated as a function of position and time for maintenance planning. High resolution results can be obtained using the mesh-based Rigorous Two-Step (R2S) method. A neutron transport simulation provides mesh-based neutron fluxes. Using these fluxes and an appropriate irradiation scenario, a nuclear inventory analysis code is used to determine the photon source density within each mesh volume element at any decay time after shutdown. The photon source density mesh is used as a source for photon transport. Photon fluxes are tallied at positions of interest and flux-to-dose rate conversion factors are used to estimate the biological dose rate. Previous work by the authors involved the creation of R2S-ACT, an open source R2S workflow, using the Direct Accelerated Geometry Monte Carlo (DAGMC) version of MCNP5 (DAG-MCNP5) for both transport steps and the ALARA nuclear inventory analysis code. Various use cases may require different combinations of transport and activation codes, and nearly all of the capabilities of R2S-ACT have broader utility. For these reasons, the individual components of R2S-ACT were reimplemented within PyNE. PyNE is a collaborative, trans-institutional, open source project consisting of Python, Cython, and C++ code pertinent to nuclear simulations and analysis. This work involved additions to the PyNE mcnp, mesh, material, alara, dagmc, and sampling modules. A lightweight r2s module was then added to connect the code from these modules. The r2s and sampling modules, together with DAG-MCNP5 and ALARA form a complete R2S workflow for obtaining shutdown dose rate results. Mesh operations, geometry discretization, physics code input/output parsing, and mesh-based source sampling capabilities have been added to PyNE generically, eliminating the need to duplicate this effort across multiple applications. The PyNE r2s module uses these capabilities to couple MCNP and ALARA to form an R2S workflow, which has been validated with the FNG shutdown dose rate benchmark. Future work will involve benchmarking the workflow using tetrahedral mesh and also non-analog sampling options. added so that they can be used interchangeably. This will allow for comparisons between results obtained by different physics codes with minimal human effort. Work will also be done to create and implement an importance sampling method for biasing neutron transport towards regions of phase space of high importance to photon dose rate tallies. This will improve the efficiency of obtaining shutdown dose rate results in large, highly-attenuating geometries. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Joint neutron measurements on the flat-top HEU critical assembly machine - 14388

Flat-Top is a critical assembly with a spherical core (either ∼17 kg of HEU or ∼6 kg of δ-phase Pu) surrounded by a spherical 1000 kg natural uranium reflector. In 2013, joint LANL and CEA measurements were conducted to measure Flat-Top configurations with the HEU core using He-3 detector systems. Similar experiments were previously documented in which the Caliban and Godiva critical assemblies were measured. The core and reflector of Flat-Top have a glory hole in which nuclear material or non-nuclear material may be placed. Five configurations were measured and included subcritical configurations at various reactivity states as well as a configuration at delayed critical. The various reactivity states were achieved by varying the placement of the safety blocks and control rods (composed of natural uranium). A description of the assembly and the specific configurations measured will be presented. The subcritical and delayed critical measurements were analyzed using the Hage-Cifarelli formalism of the Feynman Variance-to-Mean method. This analysis can be used to calculate the multiplication of each configuration, from which the multiplication factor and reactivity are determined. These results are compared to the reactivity states based upon control rod worth curves and MCNP simulations. For this work, a new MCNP model was made to accurately represent the measured configurations. The ICSBEP handbook contains a benchmark of Flat-Top but all models are very simple and do not have the required fidelity to accurately represent the measured configurations. This work also examines the number of neutrons emitted per fission for this assembly (which has a significant contribution of induced fissions in U-238 due to the large natural uranium reflector). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗