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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

A Methodology for Calculating Radiation Signatures

A rigorous formalism is presented for calculating radiation signatures from both Special Nuclear Material (SNM) as well as radiological sources. The use of MCNP6 in conjunction with CINDER/ORIGEN is described to allow for the determination of both neutron and photon leakages from objects of interest. In addition, a description of the use of MCNP6 to properly model the background neutron and photon sources is also presented. Examinations of the physics issues encountered in the modeling are investigated so as to allow for guidance in the user discerning the relevant physics to incorporate into general radiation signature calculations. Furthermore, examples are provided to assist in delineating the pertinent physics that must be accounted for. Finally, examples of detector modeling utilizing MCNP are provided along with a discussion on the generation of Receiver Operating Curves, which are the suggested means by which to determine detectability radiation signatures emanating from objects.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

HFIRCON Version 1.0.5 User Guide

The High Flux Isotope Reactor (HFIR) Controller (HFIRCON) code is a collection of python routines and C plugins that automate the workflow for fuel and single- or multicycle target depletion analyses for HFIR at Oak Ridge National Laboratory (ORNL). This code calls the LAVAMINT (LAVA Model Interrogator) parallel (MCNP) Monte Carlo N-Particle model interrogator to stochastically calculate cell volumes and bounding boxes, the ADVANTG (Automatic Variance Reduction Generation) code package for all variance reduction and source biasing calculations, the ORNL-Transformative Neutronics/MCNP5 transport solver for all transport solutions, and the MSX_DEPLETE module to perform all depletion calculations via the ORIGEN (Oak Ridge Isotope Generation) application programming interface. It also performs a robust set of postprocessing functions to automatically provide summaries of several key metrics that are common to a wide variety of typical HFIR design and safety-basis analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MCNP Simulation of the SFRC and Minimum Detection Threshold Determination (Rev. 1)

In this report the Minimum Detection Threshold (MDT) for the Spent Fuel Rod Counter (SFRC) was determined using Monte Carlo N-Particle transport computer code (MCNP). A model of the SFRC was simulated in addition to a fuel transport pipe and spent Magnox fuel at descending positions down the pipe. The neutron MDT for Magnox fuel free-falling along a discharge pipe was calculated to be equal to 4080 n/s. The gamma MDT for the same free falling fuel as measured by the SFRC’s ionization chamber was determined to be 455 γ/s. These values were determined to be a conservative value to use when the instrument is deployed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Concrete Compositions Used in Neutronics Analyses for Design of the Second Target Station Project

The purpose of this letter is to document the concrete compositions used by the Second Target Station (STS) Project Neutronics Group. Members of the Neutronics Group all have access to these compositions in our MCNP Master Model, which is stored on the ORNL GitLab server. It is expected that all STS Neutronics staff members will use these compositions in their analyses, unless all stakeholders agree to an exception. It is impossible to know the detailed composition of concrete before its constituents have been purchased, mixed, and a chemical analysis performed. Therefore, the STS Neutronics Group decided to use the same concrete compositions used by the Neutronics Group at the First Target Station (FTS). Furthermore, this leads to consistency between the analyses performed by the FTS and STS Neutronics Groups. The most important parameter to match between design simulations and construction regarding these concrete compositions is the density. Next in importance is the fraction of iron and hydrogen in the mixture.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Benchmarking a Defeatured Geant4 NIF Model with HTOAD and HNED Neutron Data

This study benchmarks a simplified Geant4 NIF Target Chamber (TC) model against foil measurements in two instruments: the HTOAD and HNED Snout. The number of product atoms per source neutron per gram ( N 0 /n/g ) is compared across multiple locations and material configurations. The model reproduces high energy threshold reactions dominated by 14 MeV neutrons to within a few percent, depending on location. Discrepancies occur where scattering and moderation are significant for low energy threshold reactions. The defeatured Geant4 model achieves comparable statistical precision with ∼400 times less CPU time than a full fidelity TC model in MCNP. The results can be produced locally on a laptop without the need for high performance computing. Low energy biasing can be improved by preserving room return pathways and implementing IRDFF cross section evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Python was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.

depletion↗
DOE OSTI2018

Dose Uncertainty Quantification in AGC Experiment

This memorandum formally documents completion of the Advanced Reactor Technologies Level 3 Milestone (M3NT-18IN060504037), “Interoffice Memorandum detailing status of the AGC irradiation uncertainty analysis” due SeptemberAugust 30, 2018. This memorandum details the progress made toward the issuance of a final report detailing the AGC uncertainty analysis including the proposed methodology, progress on activities, and explanation of how the data will be used. Radiation damage estimation is an important component of the post-irradiation analysis of the AGC experiments. It depends primarily on the fast fluence, which is determined using well-established methods of spectral adjustments. These are based on best estimates from models (e.g. MCNP). Input cross-sections, and measured activities from flux wires in the experiment, propagate uncertainties in each of these. While the methods of propagating uncertainty are well-established, the final uncertainty estimates they provide are only as good as the estimates of uncertainty in the inputs on which they are based. The purpose of this work is to implement a method that eliminates some deficiencies in the ways input uncertainties are presently estimated. The rRadiation damage estimates for the AGC experiments are ostensibly based on experimental data in the form of flux wire activation measurements; each AGC experiment has been instrumented with flux wires at various locations along the axial length to provide this data. However, the estimatesy also depend on the results of MCNP calculations. To understand why, consider that the objective of the analysis is to estimate some number of parameters (the energy group fluences ?_i) that best reproduce the observations (the measured activities of the flux wires a_i).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

3D neutronic/thermal-hydraulic coupled analysis of MYRRHA

The current tendency in multiphysics calculations applied to reactor physics is the use of already validated computer codes, coupled by means of an iterative approach. In this paper such an approach is explained concerning neutronics and thermal-hydraulics coupled analysis with MCNPX and COBRA-IV codes using a driver program and file exchange between codes. MCNPX provides the neutronic analysis of heterogeneous nuclear systems, both in critical and subcritical states, while COBRA-IV is a subchannel code that can be used for rod bundles or core thermal-hydraulics analysis. In our model, the MCNP temperature dependence of nuclear data is handled via pseudo-material approach, mixing pre-generated cross section data set to obtain the material with the desired cross section temperature. On the other hand, COBRA-IV has been updated to allow for the simulation of liquid metal cooled reactors. The coupled computational tool can be applied to any geometry and coolant, as it is the case of single fuel assembly, at pin-by-pin level, or full core simulation with the average pin of each fuel-assembly. The coupling tool has been applied to the critical core layout of the SCK-CEN MYRRHA concept, an experimental LBE cooled fast reactor presently in engineering design stage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Performance Evaluation of an Intelligent Sensor Platform for Radiation Monitoring Applications

Accurate, rugged, and reliable radiation detection systems are important for area and environmental monitoring applications. The desire for spectroscopic capability has increased in monitoring aspects of the nuclear fuel cycle to provide fast characterization of the radiation profile of a situation, such as the planned or unplanned release of material. The reduction or elimination of having to conduct sampling for laboratory analysis can result in significant cost savings for an industry, government agency, or regulatory body. A new system, comprised of a NaI:Tl scintillator and a G-M tube, has been designed and tested, taking into account the following end-user requirements: ease-of-use, capability to network and supervise multiple units, compact form factor, low power consumption, versatility, and stability. The detector sizes were selected to accommodate a dose rate up to 1 Sv/hr. Several algorithms and analysis routines have been developed to incorporate these key needs without sacrificing on accuracy, dynamic range, nuclide identification, and sensitivity. This presentation will introduce the major hardware and software components of the platform, as well as the user interface and data analysis workflow. Key features of the hardware include an environmentally robust housing, low power signal processing electronics, patented LED-based gain stabilization, and an embedded processor for unattended instrument management and data analysis. New and improved algorithms for determining scintillator gamma dose rates, total integrated dose, and nuclide identification will be introduced. The two detector elements were modeled with MCNP and validated experimentally. The results of the radiological testing shall be presented including energy resolution, throughput, dose response, and minimum detectable activities. The dose response has been evaluated in simulations and with measurements to ensure accurate response with respect to energy spectrum of the dose field and direction of the incident radiation. Spectroscopy evaluations, such as nuclide detection probability and false alarm rates, were performed and comparisons to existing algorithms shall also be presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Towards application of a neutron cross-section uncertainty propagation capability in the criticality safety methodology - 14434

At the Paul Scherrer Institute (PSI) the methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practice, on establishing an upper subcriticality limit (USL) for the effective neutron multiplication factor (k{sub eff}), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k{sub eff} values. Recently, a new computational tool referred to as NUSS (Nuclear data Uncertainty Stochastic Sampling) was developed in order to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries in ACE format. Thereby, this tool should now allow for a quantification of ND data uncertainties in CSE and to introduce these into the CSE criteria definition. This paper concerns an application of the NUSS methodology for criticality validation studies. Calculated k{sub eff} uncertainty estimations were obtained for the current PSI validation suite including in total 149 critical benchmark configurations for which in-house MCNP(X) models were developed, covering both uranium and mixed uranium-plutonium oxide fuels. The MCNPX-2.7.0 code was used together with the ENDF/B-VII.1 ND library while the ND variance-covariance matrixes (VCM) were taken from both SCALE-6.0 44-group libraries and also from ENDF/B-VII.1 files. Based on the newly obtained information on the uncertainties, a re-assessment of the lower tolerance bound (LTB) for k{sub eff} using the total uncertainties as weighting factors is proposed as means for the present PSI CSE methodology upgrade and some possible directions of further methodology enhancements are discussed next. (authors)

42 ENGINEERING↗
DOE OSTI2016

Recovering New Type of Sources by Off-Site Source Recovery Project for WIPP Disposal - 16604

OSRP uses two types of eligible packaging configurations for Cm-244 and Cf-252 disposal to WIPP: an S-100 drum and a 12'' Pipe Overpack (POC) drum. Both types of packaging are WIPP accepted configurations for Contact Handled (CH) waste, which offer different advantages and disadvantages when used in the case of the two isotopes. The S-100 is a poly-shielded type A, 55 gallon drum, containing a 6'' diameter POC. The 12'' POC drum is a type A, 55 gallon drum without the internal poly-shield and with a pipe component of 12'' diameter. The S-100 drum presents the advantage of poly-shielding which is very important for neutron sources, such as the Cf-252, or Cm-244Be, at the expense of reduced internal volume. The 12'' POC drum presents a larger pipe component accommodating many more sources, like Cm-244 alpha emitters or shielded Cf-252, for which the CH dose rate is not an issue. All the sources prepared for disposal to WIPP are required to be special form, as defined in 49 CFR 173.403. When OSRP recovers normal form sources, these can be packaged in a field closable DOT certified Special Form Capsule (SFC) model II or model III, thus becoming special form. These capsules are then packaged in the type-A containers. The difference between the two types of packaging is that the 12'' POC can hold more sources, or a mix of sources and capsules at the expense of less shielding. This becomes important when the TRU content criteria of 100 nCi/g of TRU waste has to be met for the drums that are packaged for WIPP disposal. Measured neutron dose rates for Cm-244 packaged in S-100 agree reasonably well with calculated dose rates based on the models developed with MCNP. Future work will include validating the model for Cm-244 packaged in 12'' POC. A 1:1 ratio of neutron to gamma is expected for Cm-244Be dose rates packaged in S-100 drums. For Cf-252 packaged in S-100 and 12'' POC, the contribution of Cf-250 to the total emission neutron rate had to be taken into account due to the increasing role of Cf-250 as sources age. The uncertainty in the measurement of Cf-252 neutron dose rate for S-100 drum was quite high due to a high neutron background. A larger Cf- 252 source needs to be used for measurements in the future, in order to fully validate the model developed. A better agreement for measured versus calculated rate was obtained for a Cf-252 packaged in 12'' POC. For the WIPP disposal of these sources, OSRP will use the best packaging configuration for these isotopes in order to minimize the dose rate and maximize the TRU content. The S-100 packaging will be used for Cf-252 and Cm-244Be sources, when neutron dose reduction by the poly-shielding is important. Neutron dose rates will be estimated using the new MCNP correlations.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Aqueous Solution Vessel Thermal Model Development II

The work presented in this report is a continuation of the work described in the May 2015 report, “Aqueous Solution Vessel Thermal Model Development”. This computational fluid dynamics (CFD) model aims to predict the temperature and bubble volume fraction in an aqueous solution of uranium. These values affect the reactivity of the fissile solution, so it is important to be able to calculate them and determine their effects on the reaction. Part A of this report describes some of the parameter comparisons performed on the CFD model using Fluent. Part B describes the coupling of the Fluent model with a Monte-Carlo N-Particle (MCNP) neutron transport model. The fuel tank geometry is the same as it was in the May 2015 report, annular with a thickness-to-height ratio of 0.16. An accelerator-driven neutron source provides the excitation for the reaction, and internal and external water cooling channels remove the heat. The model used in this work incorporates the Eulerian multiphase model with lift, wall lubrication, turbulent dispersion and turbulence interaction. The buoyancy-driven flow is modeled using the Boussinesq approximation, and the flow turbulence is determined using the k-ω Shear-Stress-Transport (SST) model. The dispersed turbulence multiphase model is employed to capture the multiphase turbulence effects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Simulations of Neutron Time-of-Flight Method by Inelastic Scattering Carbon-12 using MCNP6 and Geant4

A neutron spectroscopy technique using neutron time-of-flight (TOF) with gamma tagging from inelastic scattering on carbon-12 was examined using MCNP6 and Geant4. Experimental data for the 12 C(n,n'$γ$) reaction from D. Spaargaren and C.C. Jonken was used to supplement the simulations due to their limitations with modeling correlated particles on an event-by-event basis. The MCNP and Geant4 models relied on different methods to determine the efficiencies and were compared against each other for validation of the results. The total efficiency of coincident detection as determined by MCNP is around 6.32E-11. That is, around 6 events will be detected per 100 billion source neutrons. It was found that the amount of neutrons which lose energy to elastic scattering is not negligible, and efforts to reconstruct the beam energy from the detected neutrons must take this into account. Running on 40 computer cores in multi-threading mode, the Geant4 model could simulate about half a billion source neutrons per hour. In the absence of biasing methods, Geant4 can not simulate the efficiency in any reasonable amount of time given current computing power. Further efforts will be made to introduce biasing to the simulation to verify the MCNP results.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

SU-E-T-548: Modeling of Breast IORT Using the Xoft 50 KV Brachytherapy Source and 316L Steel Rigid Shield

Purpose: Xoft provides a set of 316L Stainless Steel Rigid Shields to be used with their 50 kV X-ray source for Breast IORT treatments. Modeling the different shield sizes in MCNP provides information to help make clinical decisions for selecting the appropriate shield size. Methods: The Xoft Axxent 50 kV Electronic Brachytherapy System has several applications in radiation therapy, one of which is treating cancer of the breast intraoperatively by placing the miniaturized X-ray tube inside an applicator balloon that is expanded to fill the lumpectomy bed immediately following tumor removal. The ribs, lung, and muscular chest wall are all regions at risk to receive undesired dose during the treatment. A Xoft 316L Stainless Steel Rigid Shield can be placed between the intracostal muscles of the chest wall and the remaining breast tissue near the balloon to attenuate the beam and protect these organs. These shields are provided in 5 different sizes, and the effects on dose to the surrounding tissues vary with shield size. MCNP was used to model this environment and tally dose rate to certain regions of interest. Results: The average rib dose rate calculated using 0cm (i.e., no shield), 3cm, and 5cm diameter shields were 26.89, 15.43, and 8.91 Gy/hr respectively. The maximum dose rates within the rib reached 94.74 Gy/hr, 53.56 Gy/hr, and 31.44 Gy/hr for the 0cm, 3cm, and 5cm cases respectively. The shadowing effect caused by the steel shields was seen in the 3-D meshes and line profiles. Conclusion: This model predicts a higher dose rate to the underlying rib region with the 3cm shield compared to the 5cm shield; it may be useful to select the largest possible diameter when choosing a shield size for a particular IORT patient. The ability to attenuate the beam to reduce rib dose was also confirmed. Research sponsored by Xoft Inc, a subsidiary of iCAD.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2016

Neutron flux and power in RTP core-15

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core with literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗