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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
NASA NTRS2014

Path Toward a Unifid Geometry for Radiation Transport

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex computer-aided design (CAD) models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN [high charge and energy transport code developed by NASA Langley Research Center (LaRC)], are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading to the creation and maintenance of toolkit-specific simplistic geometry models. The work presented here builds on the Direct Accelerated Geometry Monte Carlo (DAGMC) toolkit developed for use with the Monte Carlo N-Particle (MCNP) transport code. The workflow for achieving radiation transport on CAD models using MCNP and FLUKA has been demonstrated and the results of analyses on realistic spacecraft/habitats will be presented. Future work is planned that will further automate this process and enable the use of multiple radiation transport codes on identical geometry models imported from CAD. This effort will enhance the modeling tools used by NASA to accurately evaluate the astronaut space radiation risk and accurately determine the protection provided by as-designed exploration mission vehicles and habitats

Lee, Kerry↗
DOE OSTI2022

Chemical mixture exposure patterns and obesity among U.S. adults in NHANES 2005–2012

The effect of chemical exposure on obesity has raised great concerns. Real-world chemical exposure always imposes mixture impacts, however their exposure patterns and the corresponding associations with obesity have not been fully evaluated. To discover obesity-related mixed chemical exposure patterns in the general U.S. population. Sparse Decompositional Regression (SDR), a model adapted from sparse representation learning technique, was developed to identify exposure patterns of chemical mixtures with exclusion (non-targeted model) and inclusion (targeted model) of health outcomes. We assessed the relationships between the identified chemical mixture patterns and obesity-related indexes. We also conducted a comprehensive evaluation of this SDR model by comparing to the existing models, including generalized linear regression model (GLM), principal component analysis (PCA), and Bayesian kernel machine regression (BKMR). Eight core exposure patterns were identified using the non-targeted SDR model. Patterns of high levels of MEP, high levels of naphthalene metabolites (ΣOH-Nap), and a pattern of high exposure levels of MCOP, MCNP, and MCPP were positively associated with obesity. Patterns of high levels of BP3, and a pattern of higher mixed levels of MPB, PPB, and MEP were found to have negative associations. Associations were strengthened using the targeted SDR model. In the single chemical analysis by GLM, BP3, MBP, PPB, MCOP, and MCNP showed significant associations with obesity or body indexes. The SDR model exceeded the performance of PCA in pattern identification. Both SDR and BKMR identified a positive contribution of ΣOH-Nap and MCOP, as well as a negative contribution of BP3 and PPB to obesity. Our study identified five core exposure patterns of chemical mixtures significantly associated with obesity using the newly developed SDR model. The SDR model could open a new avenue for assessing health effects of environmental mixture contaminants.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Comparison of the Performance of Various Correlated Fission Multiplicity Monte Carlo Codes

When measuring special nuclear material for a large array of national security needs, it is often important to accurately predict outcomes of the physical experiment using Monte Carlo radiation transport codes. Historically, radiation transport codes have uncorrelated fission emissions. In reality, both spontaneous and induced fissions release particles that are correlated in time, energy, and multiplicity. This work investigates the performance of various current Monte Carlo codes that take into account the correlated physics of fission neutrons. Only fission neutrons are of interest and the physics of gamma production in fission is ignored. Because of their large impact on correlated neutron results, underlying fission neutron multiplicity distributions utilized by the different codes are also compared. The codes currently being compared include MCNP{sup R}6, MCNP{sup R}6/FREYA, and MCNPX-PoliMi. By default, MCNP{sup R}6 uses a bounded integer treatment to sample the number of neutrons emitted from each fission event. The FMULT card, an optional input in MCNP that allows for user definition of spontaneous and induced fission parameters, can be utilized to call either built-in or user-specified multiplicities to replace the bounded integer treatment. Similarly, MCNPX-PoliMi utilizes one of a few different built-in multiplicity sets. MCNPX-PoliMi also models the multiplicity dependence of emitted neutron energy in spontaneous fission. It is possible to use an anisotropic model for emitted fission neutrons in MCNPXPoliMi, but this option was not used. For MCNP{sup R}6/FREYA, the FREYA fission event generator determines the number of particles emitted for each fission event and gives the results to MCNP{sup R}6 for transport. The fission event generator uses fission fragment mass and kinetic energy distributions, unbounded statistical evaporation models, and conservation of energy and momentum to generate the number, energy, and direction of neutrons released by each fission event. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Eigenfunction Decomposition of Reactor Perturbations and Transitions Using MCNP Monte Carlo

Monte Carlo criticality calculations for nuclear reactors have typically only been able to find the primary eigenvalue and eigenmode. With the new fission matrix capability in MCNP6, it becomes possible to solve for higher modes, which are useful in wide variety of analyses. As a test case of this new capability, two reactor models in both a base configuration and a perturbed configuration were simulated in MCNP to generate fission matrices. Tools were written to find a relatively small number of eigenpairs of the resulting matrices, whereupon it was found that the implicitly restarted Arnoldi method outperformed the previously used power iteration as an eigenvalue algorithm by a factor of 1500 for a 3600 × 3600 sparse matrix.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Experimental validation of a nuclear forensics methodology for source reactor-type discrimination of chemically separated plutonium

An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. Finally, the analyses validate the nuclear forensics methodology developed.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

sUAS Support for Mars 2020

The release of airborne radioactive material presents a health risk hazard to many individuals, including emergency responders and the public. Characterizing the unknown radioactive dangers produced during such incidents is essential. An advantage to utilizing unmanned aerial systems in this effort is the reduction of risk to personnel otherwise tasked with characterization, as well as more quickly obtaining data points during cloud immersion. By pairing a continuous air monitor (CAM) and detector with a drone, quantification of unknown material in the plume can be determined, and thus the extent of dose to people in the area can be estimated. A small, low-flow air sampler with a Geiger-Mueller counter may be characterized using experimental measurements of point sources representing a cloud of material and modelled using the Monte Carlo N-Particle Transport Code (MCNP), a software package for simulating nuclear processes, to find the usefulness and advantages of the system for detection of nuclides of concern in various nuclear plume scenarios. The flying time of small-unmanned aircraft systems (sUASs) is limited; thus, the minimum detectable activity and concentration rates possible must be determined over the system’s window of operation. While the activity concentrations corresponding to external and internal doses of concern can be detected with certainty for 90 Sr and 137 Cs in a plume, those corresponding to 238 Pu are more difficult to detect. The central principle of employing the sUAS is that the cockpit does not have an operator onboard the aircraft; therefore, control of the aircraft must take place by other means. With a sUAS, command and control can be separated into three distinct methods: (1) ground control or remote piloting, (2) semi-autonomous control, and (3) autonomous control. The isotope 238 Pu is used in the space industry, providing long-lived heat generation for fuel in outer space (NASA 2014). Due to the inherent risks with rocket launches to get objects into space, monitoring for release of these hazardous substances is a major undertaking. The current means is through employment of environmental CAMs. It is conceivable, however, that drone monitoring for release of 238 Pu in the event of a space launch mishap would give a greater chance of detection of plutonium particulate, because the particulate’s diffusion in air upon settling to the ground would make the probability of collecting a particle on a filter in a CAM very low.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

SRNL Development of Recovery Processes for Mark-18A Heavy Actinide Targets

Savannah River National Laboratory (SRNL) and Oak Ridge National Laboratory (ORNL) are developing plans for the recovery of rare and unique isotopes contained within heavy-actinide target assemblies, specifically the Mark-18A. Mark-18A assemblies were irradiated in Savannah River Site (SRS) reactors in the 1970s under extremely high neutron-flux conditions and produced, virtually, the world's supply of plutonium-244, an isotope of key importance to high-precision actinide measurement and other scientific and nonproliferation uses; and curium highly enriched in heavy isotopes (e.g., curium-246 and curium-248). In 2015 and 2016, SRNL is pursuing tasks that would reduce program risk and budget requirements, including further characterization of unprocessed targets; engineering studies for the use of the SRNL Shielded Cells Facility (SCF) for recovery; and development of onsite and offsite shipping methods including a replacement for the heavy (70 ton) cask previously used for onsite transfer of irradiated items at SRS. A status update is provided for the characterization, including modeling using the Monte Carlo N-Particle Transport Code (MCNP); direct non-destructive assay measurements; and cask design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Experimental Steps toward a Density Law for Chlorine-Crediting Criticality Models of Aqueous Plutonium solutions

Criticality accident prevention is an essential safety consideration for all operations involving fissionable and fissile materials. Predictive criticality calculations using advanced neutron transport codes, such as Monte Carlo N-Particle Transport (MCNP), are invaluable tools for designing and implementing operational limits. While indispensable, these tools are limited by the quality and accuracy of the inputs that the user provides to define the modeled system. Parameters such as atomic composition, shape, and material density must be accurately defined to obtain a meaningful result. In the case of material property details for a plutonium model, accurate material characterization data is sometimes sparse. Thus limitations in calculation accuracy can be reduced by improving our knowledge of material properties in the targeted fissile systems. One of the biggest remaining challenges to accurately defining fissile systems is the description of aqueous fissile solutions. Even simple properties, such as density, are not well-known for fissile solutions relevant to nuclear energy and security. Described here are initial efforts undertaken to improve criticality calculation inputs for fissile plutonium chloride solutions in water. This effort is focused on experimentally determining accurate solution characteristics for ternary plutonium chloride/hydrochloric acid/water systems, by measurement of water activity and solution density. The effect of inputting experimental densities for these solutions into MCNP criticality calculations is compared to the traditional approach of modeling an idealized (and fictitious) plutonium-water mixture. Expansion of these efforts to a working density law for aqueous plutonium chloride solutions is also discussed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Initial Investigation of Coupled Electron-Photon Transport with SCEPTRE and RAPTURE (Rev.1.1)

The SCEPTRE1 code models coupled photon and electron transport in material media using discrete ordinates methods. The code can use different solvers in 1D and 3D in parallel, promising users the ability to run problems that are significant both in physics and computational scale. One capability that is perhaps unique at this time in the NNSA complex is SCEPTRE’s coupled photon– electron transport physics in a discrete ordinates approach. For example, incident x–rays can generate a significant electron population, whose transport can be simulated within the same code. In this context, fully–coupled means that photons can produce electrons and electrons can produce photons; uncoupled or no–coupling means photons do not produce electrons. Partial–coupling, an intermediate option in which photons can produce electrons but electrons do not produce photons, is not considered in the present work. The MCNP code can treat this physics with a Monte Carlo approach. MCNP should be viewed as complementary to SCEPTRE in both computational method (Monte Carlo versus discrete ordinates) as well as data libraries. SCEPTRE and MCNP should provide important checks on each other, especially from complementary modeling frameworks of a S n (discrete ordinates) mesh based method versus Monte Carlo particle based method. In order to learn more about SCEPTRE, and how it might inform engineering work at the Laboratory, we used it to carry out a set of initial studies. This was a series of 1D simulations of photon beams hitting Aluminum slabs. Our studies mainly used SCEPTRE directly, and made little use of the closely associated RAPTURE code (more on this below). The goals of these studies were first, to familiarize ourselves with operating the code, and second, to begin to assess the importance of coupled electron–photon transport. To support these studies, we began to develop a set of scripts to initialize inputs, run tools, process outputs, and plot results. An initial version of these scripts has been committed to the Lightning workflow repository. This memo relates the results of those studies.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Neutron Yield of Thermo Scientific P385 D-T Neutron Generator vs . Current and Voltage

The Thermo Scientific P385 Neutron Generator is a compact neutron source, producing 14 MeV neutrons through the deuterium-tritium (DT) fusion reaction. It is important to measure and understand the dependence of the neutron production rate on the accelerator current and voltage. In this study we evaluated neutron production with an absolutely calibrated liquid scintillator neutron spectrometer (BTI N-Probe), an absolutely calibrated He-3 detector surrounded by HDPE shells (Detec Nested Neutron Spectrometer, NNS), and two uncalibrated ZnS fast neutron scintillators (EJ-410), for both A3082 and A3083 sealed tubes. Here we also modeled the neutron yield using the TRIM code, which calculates the trajectory and the energy loss of deuterons and tritons within the target. Experimental results showed an essentially linear dependence on beam current, as expected. A 3.59 ±0.08 power law dependence on the operating voltage was measured, in effective agreement with the modeled value of 3.5. A series of absolute NNS and N-Probe measurements, matched against MCNP calculations, showed that the A3083 and A3082 tubes provide a maximum neutron yield of 8.2 × 10 8 n/s and 4.7 × 10 8 n/s respectively, with estimated uncertainty of ±10%.We showed, through modeling, that tritium decay is not a significant consideration for tubes, such as these, with lifetimes of less than 10 years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Critical Assemblies: Dragon Burst Assembly and Solution Assemblies

This work reviews the historical literature associated with the Dragon experiment and water boiler reactors operated at Los Alamos during the Manhattan Project. Frisch’s invited talk given at the American Nuclear Society’s Fast Burst Reactor Conference held at the University of New Mexico in Albuquerque, New Mexico, in 1969 is quoted. From the literature review, basic models for the Dragon experiment and for a water boiler–type assembly (aqueous homogeneous reactor) were created that can be used for conducting multiphysics simulations for criticality excursion studies. This methodology utilizes the coupled neutronic-hydrodynamic method to perform a time-dependent dynamic simulation of a criticality excursion. MCNP® was utilized to calculate important nuclear kinetic parameters that were incorporated into the models. Simulation results compare reasonably well with historic data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Resolved Resonance Region Analysis of 206 Pb, 207 Pb, and 208 Pb for Next Generation Lead-Cooled Fast Systems

Recently, great interest has been generated in using lead as a coolant for fast neutron systems and as a result it is important to investigate the ENDF/B-VIII.0 isotopic evaluations that comprise stable le ad. To this end, resonance parameters for 206 Pb, 207 Pb, and 208 Pb were re-evaluated because their resolved reso­nance regions extend beyond 0.5 MeV meaning resonance parameters used in reconstructing cross sections and elastic scattering angular distributions impact fast systems. The impact of resonance parameters is demon­strated by the differences between the evaluations in predicting experimental results from the fully modeled RPI Quasi-Differential Scattering Experiment via MCNPv6.2. In addition, MCNP KCODE calculations of lead-sensitive fast spectra critical benchmarks showed variations of keff on average of 400 pcm, caused solely from differences in elastic scattering angular distributions in 208 Pb. Re-evaluation entailed fitting data with the program SAMMY.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Trabecula: A Random Generator of Computational Phantoms for Bone Marrow Dosimetry

This study was motivated by the efforts on evaluation of radiation risk for leukemia incidence in the Techa River Cohort, where the main bone marrow dose contributors were 89,90Sr (bone-seeking beta emitters). Energy deposition in bone marrow targets was evaluated by simulating radiation particle transport using computational phantoms. The present paper describes the computer program “Trabecula” implementing an algorithm for parametric generation of computational phantoms, which serve as the basis for calculating bone marrow doses. “Trabecula” is a user-friendly tool that automatically converts analytical models into voxelized representations that are directly compatible as input to MCNP.

Shishkina, Elena A.↗
DOE OSTI2015

SU-E-T-297: Dosimetric Assessment of An Air-Filled Balloon Applicator in HDR Vaginal Cuff Brachytherapy Using the Monte Carlo Method

Purpose: As an alternative to cylindrical applicators, air inflated balloon applicators have been introduced into HDR vaginal cuff brachytherapy treatment to achieve sufficient dose to vagina mucosa as well as to spare rectum and bladder. In general, TG43 formulae based treatment planning systems do not take into account tissue inhomogeneity, and air in the balloon applicator can cause higher delivered dose to mucosa than treatment plan reported. We investigated dosimetric effect of air in balloon applicator using the Monte Carlo method. Methods: The thirteen-catheter Capri applicator with a Nucletron Ir-192 seed was modeled for various balloon diameters (2cm to 3.5cm) using the MCNP Monte Carlo code. Ir-192 seed was placed in both central and peripheral catheters to replicate real patient situations. Existence of charged particle equilibrium (CPE) with air balloon was evaluated by comparing kerma and dose at various distances (1mm to 70mm) from surface of air-filled applicator. Also mucosa dose by an air-filled applicator was compared with by a water-filled applicator to evaluate dosimetry accuracy of planning system without tissue inhomogeneity correction. Results: Beyond 1mm from air/tissue interface, the difference between kerma and dose was within 2%. CPE (or transient CPE) condition was deemed existent, and in this region no electron transport was necessary in Monte Carlo simulations. At 1mm or less, the deviation of dose from kerma became more apparent. Increase of dose to mucosa depended on diameter of air balloon. The increment of dose to mucosa was 2.5% and 4.3% on average for 2cm and 3.5cm applicators, respectively. Conclusion: After introduction of air balloon applicator, CPE fails only at the proximity of air/tissue interface. Although dose to mucosa is increased, there is no significant dosimetric difference (<5%) between air and water filled applicators. Tissue inhomogeneity correction is not necessary for air-filled applicators.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

Creating NDA working standards through high-fidelity spent fuel modeling

The Next Generation Safeguards Initiative (NGSI) is developing advanced non-destructive assay (NDA) techniques for spent nuclear fuel assemblies to advance the state-of-the-art in safeguards measurements. These measurements aim beyond the capabilities of existing methods to include the evaluation of plutonium and fissile material inventory, independent of operator declarations. Testing and evaluation of advanced NDA performance will require reference assemblies with well-characterized compositions to serve as working standards against which the NDA methods can be benchmarked and for uncertainty quantification. To support the development of standards for the NGSI spent fuel NDA project, high-fidelity modeling of irradiated fuel assemblies is being performed to characterize fuel compositions and radiation emission data. The assembly depletion simulations apply detailed operating history information and core simulation data as it is available to perform high fidelity axial and pin-by-pin fuel characterization for more than 1600 nuclides. The resulting pin-by-pin isotopic inventories are used to optimize the NDA measurements and provide information necessary to unfold and interpret the measurement data, e.g., passive gamma emitters, neutron emitters, neutron absorbers, and fissile content. A key requirement of this study is the analysis of uncertainties associated with the calculated compositions and signatures for the standard assemblies; uncertainties introduced by the calculation methods, nuclear data, and operating information. An integral part of this assessment involves the application of experimental data from destructive radiochemical assay to assess the uncertainty and bias in computed inventories, the impact of parameters such as assembly burnup gradients and burnable poisons, and the influence of neighboring assemblies on periphery rods. This paper will present the results of high fidelity assembly depletion modeling and uncertainty analysis from independent calculations performed using SCALE and MCNP. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of a high-fidelity multi-physics coupling between MCNP6.2 and CTF4.0 for VVER applications

Ensuring system safety in the design, licensing, and operation phases is a priority in the nuclear industry. Performing extensive, full-scale reactor safety experiments is often prohibitive due to the large associated costs. Computational simulations offer an alternative safety analysis method, typically with significant cost reductions. Recent high-level developments in technology and increased availability of computational resources have allowed the development of high-fidelity, high-resolution multi-physics coupled codes. Such developments may be used to generate reference models for deterministic core calculations. Under this framework, the high-fidelity continuous energy Monte Carlo-based neutron transport code MCNP6.2 was coupled externally with the state-of-the-art thermal-hydraulics subchannel code, CTF4.0 for VVER (Water-Water Energetic Reactor) applications. A VVER-1000 fuel assembly model was used to demonstrate the capability of the coupled code. The converged coupled solution is compared to initial results, consisting of the first MCNP evaluation and first CTF evaluation after initialization. The VVER-1000 type assembly results are compared to other evaluations of the same assembly model. The findings indicate good agreement with expectations and reference cases, where available. The initial results of the coupled MCNP6.2/CTF4.0 calculations at the assembly level for steady-state calculations are presented in this study, which may support future work toward high-fidelity coupled full-core modeling capabilities. The multi-physics model may be further improved, enhanced, and expanded for both cycle depletion and transient applications. Such a multi-physics system will also be applicable to the VVER-1200 and other triangular lattice designs and support their deployment and operation safely and economically. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Designing Flexibility into the Authorized Radioactive Contents of the 9982 Shipping Package (Midterm Report)

With regards to the planned contents for the 9982, three sets of radionuclides have been analyzed. A set of percentages by weight for plutonium from 3013 canisters, training sources sent in 9977’s, and americium standards from 9978’s were used as the bases for realistic contents that could be shipped in a 9982 package. Using the tables and limits located in G-BDR-A-00001 Rev. E, the calculations began by determining the mass limits for the radionuclides in each source. These mass limits were then used alongside the mass of the radionuclides in the sum-of-the-fractions methodology present in the content description. If this summation is less than or equal to one, the contents are allowed for shipment in the 9982. For the aforementioned sets, the listed radionuclides could be shipped as-is; however, the goal of the calculations is to determine the maximum content mass allowed. To do so, the percentages of weight were used to find where the summations would equal one rather than simply satisfy the inequality. By solving for the content mass when the inequality equals one, the maximum content mass is calculated, which can then be multiplied by each percent weight to get each radionuclide’s mass. The MCNP input files have also been compared to find differences between the oxide and metals forms of the radionuclides. While there are several differences present that make sense, such as oxides have more mass and a lower density due to the presence of oxygen in the oxide forms, there are other differences that are possibly unintentional. With the input files for the metals, both with and without a shielding pig, neutrons are being used in the photon models, seen by “imp:n” being present within the file instead of “imp:p”, where “imp” stands for the importance of a region for either neutrons or photons. Effectively, this results in the dose due to photons being modelled with neutrons instead, potentially resulting in the calculated dose rates being incorrect. As MCNP is proprietary, I do not have access to it here and must consult with the person handling the shielding calculations to discuss what these differences may affect and how to change them. If he remains out of the office or if fixing these files takes longer than expected, the final report may focus more on the contents while discussing these differences; however, it could potentially lack discussion on finalizing the planned contents due to the mass limits still needing to be calculated.

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

Characterization of HPGe gamma spectrometric detectors systems for Instrumental Neutron Activation Analysis (INAA) at the Colombian Geological Survey

This paper presents the progress made by the Neutron Activation Analysis (NAA) laboratory at the Colombian Geological Survey (SGC in its Spanish acronym), towards the characterization of its gamma spectrometric systems for Instrumental Neutron Activation Analysis (INAA), with the aim of introducing corrections to the measurements by variations in sample geometry. Characterization includes the empirical determination of the interaction point of gamma radiation inside the Germanium crystal, through the application of a linear model and the use of a fast Monte Carlo N-Particle (MCNP) software to estimate correction factors for differences in counting efficiency that arise from variations in sample density between samples and standards.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗