High Gain MCP ALD Films
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33 ADVANCED PROPULSION SYSTEMS↗
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Sample Preparation Laboratory-SPL
Presentation on the design features of the Sample Preparation Laboratory currently under review at MFC.
36 MATERIALS SCIENCE↗
OVERVIEW OF RECENT PU-238 PRODUCTION ACTIVITIES AT IDAHO NATIONAL LABORATORY
The Plutonium Fuel Services (PFS) program at Idaho National Laboratory (INL) is active in the qualification of irradiation targets containing Np-237 for irradiation in the Advanced Test Reactor (ATR) to produce Pu-238 for future NASA missions. INL qualified and loaded 7 targets in ATR’s South Flux Trap (SFT) for cycle 169A, which occurred in Spring 2021. This program was reinitiated after two baseline production targets in three positions validated significant production of Pu-238 [ref.1]. The validation model was followed by the PFS-1 experimental test in the ATRC (Critical) facility [ref.2]. This paper outlines the progress and status of the PFS program. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR will be discussed.
36 MATERIALS SCIENCE↗
Tracking and Monitoring with Dosimeter-Enabled ARG-US RFID System - 12009
Automated monitoring and tracking of materials with radio frequency identification (RFID) technology can significantly improve both the operating efficiency of radiological facilities and the application of the ALARA (as low as reasonably achievable) principle in them. One such system, called ARG-US, has been developed by Argonne National Laboratory for the U.S. Department of Energy (DOE) Packaging and Certification Program to use in managing sensitive nuclear and radioactive materials. Several ARG-US systems are in various stages of deployment and advanced testing across DOE sites. ARG-US utilizes sensors in the tags to continuously monitor the state of health of the packaging and promptly disseminates alarms to authorized users. In conjunction with global positioning system (GPS) tracking provided by TRANSCOM, the system can also monitor and track packages during transport. A compact dosimeter has been incorporated in the ARG-US tags via an onboard universal asynchronous receiver/transmitter interface. The detector has a wide measurement range for gamma radiation - from 0.1 mSv/h to 8 Sv/h. The detector is able to generate alarms for both high and low radiation and for a high cumulative dose. In a large installation, strategically located dosimeter-enabled tags can yield an accurate, real-time, 2D or 3D dose field map that can be used to enhance facility safety, security, and safeguards. This implementation can also lead to a reduced need for manned surveillance and reduced exposure of personnel to radiation, consistent with the ALARA principle at workplaces. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Panel Session 82: Shareholder Engagement with DOE Associate Undersecretary's Environment, Health, Safety, and Security
This panel discussed the mission of the US DOE Office of the Associate Under Secretary for the Office of Environment, Health, Safety and Security (AU) and its functional support to the overall mission of the DOE, DOE Office of Environmental Management, and the contractor partners that support that overall mission. Participation in this panel helps achieve one of AU's strategic plan goals - to engage stakeholders to ensure AU's Environment, Health, Safety and Security (EHS and S) policy and support activities align with DOE mission needs. The Panel included stakeholders to discuss their perspectives on AU stakeholder engagement as well as representatives from four policy and support offices from AU. This panel was an important implementation of the DOE Office of the Associate Under Secretary for the Office of EHS and S (AU) strategic plan goals - to engage stakeholders to ensure AUs Environment, Health, Safety and Security (EHS and S) policy and support activities align with DOE mission needs. The panel included stakeholders to discuss their perspectives on AU stakeholder engagement as well as representatives from four policy and support offices from AU including the Office of Health and Safety (AU-10); Office of Environmental Protection and ES and H Reporting (AU-20); Office of Nuclear Safety (AU-30); and Office of Security (AU-50). Presentations proposed include: 1 - Office of Environment, Health, Safety and Security (AU) - Stakeholder Engagement (Matthew Moury, US DOE, Office of Environment, Health, Safety and Security (USA)); 2 - Environment, Health, Safety, and Security Policy Making: Rules, Directives, Technical Standards and Programs (Patricia Worthington, US DOE, Office of Environment, Health, Safety and Security (USA)); 3 - Environment, Health, Safety, and Security: Corporate Programs (Josh Silverman, US DOE, Office of Environment, Health, Safety and Security (USA)); 4 - Environment, Health, Safety and Security: Technical Assistance (Marc Brooks, US DOE, Office of Environment, Health, Safety and Security (USA)); 5 - Environment, Health, Safety and Security: Goals and Initiatives (Tom Hiltz, US DOE, Office of Environment, Health, Safety and Security (USA)); 6 - Shareholder Engagement with DOE Associate Undersecretary's Environment, Health, Safety, and Security (Greg Sosson, US DOE (USA)); 7 - Nuclear Safety Center of Excellence (Chip Lagdon, Bechtel National, Inc (USA)); 8 - EFCOG Safety Working Group (Jan Preston, Fluor Corporation (USA))
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Using Python For 3D Visualizations of MCNP output
MCNP is ubiquitously used within industries, universities and national laboratories as a tool for accurately modeling radiation. With so much focus on the details of transport, attention is seldom given to the visualizations of these meticulously constructed outputs. Python is a programing language designed to promote code readability and has steadily grown in popularity since the 1990's. Python is also a heavily crowd-sourced programing language which generally means that the problems that you are trying to solve with computer codes are already solved; The user is often left with the now-simplified task of cobbling together a good and useful compilation of these tools and adapting them to solve the problem at hand. I have taken advantage of this situation and constructed some visualization tools in python that can be used to duplicate some of the 2D radiation fields that MCNP produces and also produce a meaningful 3D representation of a radiation field. (author)
61 RADIATION PROTECTION AND DOSIMETRY↗
Characterizing Fluence as a Function of Areal Density for 1 and 14 MeV Neutrons Transported Through Air
Computing neutron dosimetry involves transporting neutrons [n] from a source to a sink, determining the neutron heating within the sink and multiplying the neutron heating by a quality factor to compute the dose equivalent. The space and energy dependent neutron fluence is a measure of the amount of neutrons crossing a particular surface area, about a particular energy and is given in units of [n/cm{sup 2}]. Microscopic cross sections are given in units of area and vary by isotope and reaction type as a function of energy. The product of the atom density of a particular atom in a specie and the microscopic cross section of a particular reaction results in the macroscopic cross section of a particular reaction, given in units of per length [1/cm] as a function of energy. Summing the macroscopic cross sections of all isotopes, and all collisions resulting in heating, within a volume, results in a total collision heating macroscopic cross section for that volume. The product of the energy and spatial dependent collision heating macroscopic cross section and the energy and spatial dependent fluence, within an energy group, results in a collision heating reaction density for that particular reaction in that energy group. Weighting each energy group's collision heating reactions, with the corresponding energy dependent particle damage per collision heating event (usually biological damage), and then integrating over energy and volume results in the delivered dose. Therefore understanding dose delivery involves examining the spatial and energy (and time) dependent fluence behavior. Because of the coupled energy and space dependence of the heating weighting and macroscopic cross sections, the fluence is not so easily separable in energy and space; however, the energy integrated fluence can be used as a first order approximation of the relative magnitude of reactions as a function of space. Here we examine the first part in the dosimetry calculation process, transporting neutrons from source to sink, as applied to neutron transport in air, examining characteristics of the fluence. Presented here is analysis regarding the nature of the fluence peak between 10 g/cm{sup 2} and 100 g/cm{sup 2} for a typical air transport calculation. Due to the high concentration of nitrogen and oxygen isotopes of these elements, these isotopes tend to dominate the transport. From the analysis presented here, the fluence peak is the result of scatter back and forth across the peak regime, and (n,2n) has a minimal (<4%) contribution. Further evidence of minimal (n,2n) impact in the 10-100 g/cm{sup 2} span includes the 1 MeV fluence peak being so much larger than the 14 MeV peak; where 1 MeV is not above the (n,2n) threshold. Also presented here is the usefulness of using tally tagging with cell flagging to further dissect the nature of the fluence. These technologies allow us to understand the originating volume that makes a contribution to a scoring regime. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
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↗
Validation of Methodology to Simulate Gamma Doses from Historical Weapons Tests
The United States tested nuclear weapons in the atmosphere (aboveground) from 1945 until 1963, when the nation became a signatory to the Partial Test Ban Treaty. During those tests, gamma doses were often measured within kilometers of the point of detonation. Additional gamma dose measurements were made in and around the location of any personnel conducting the test, possibly with personal dosimeters. These gamma dose measurements have been declassified to provide the personnel involved with an estimate of the total dose they received while involved with the aboveground testing program. Several years ago, interest in these gamma doses was renewed by organizations that would need to provide the public with dose estimates in the event of a radiological or nuclear weapon terrorist attack on the United States. This work provides an initial validation of the computational methods used to simulate dose from a radiological or nuclear weapon. This initial work will focus on dose from gamma rays. The sources of gamma rays from a nuclear weapon include prompt fission gammas, gammas generated via neutron interactions (capture, inelastic scattering, and activation), and gammas produced by the decay of fission products. All of these gamma ray sources for a nuclear weapon are also potential sources of gammas from a radiological weapon. However, the sources of gamma rays from a radiological weapon depend on the material present in the device, and prompt fission gammas will mostly come from spontaneous fission instead of induced fission. This paper focuses on validation using gamma ray dose measurements from nuclear detonations, which is adequate to validate the requisite transport methodologies, source term generation, and cross sections needed to model gamma doses delivered by most radiological weapons. The 1961 report on Operation Plumbbob provides an overview of how the total gamma dose delivered by a nuclear weapon is broken down by the different sources specified above. The prompt fission gamma rays contribute about 5% of the total gamma dose. The gamma rays produced by neutron interactions usually deliver the largest portion of the total gamma dose, which is typically between 55% and 85% of the total. The remainder of the total gamma ray dose is attributed to decay of fission products.
61 RADIATION PROTECTION AND DOSIMETRY↗
Effects of statistical uncertainty in activation calculations based on Monte Carlo radiation transport - 25353
This paper focuses on the analysis of the uncertainty in isotopic concentrations determined by Monte Carlo depletion calculations performed for the purposes of activation source calculations. In this work, we use repeated, statistically independent, Monte Carlo depletion calculations to examine the convergence behavior and statistical uncertainty in isotopic densities. Performing many, small, Monte Carlo depletion simulations and then recombining them is a useful approach and enables statistical analysis that can help ensure convergence. Convergence behavior is discussed in terms of stability of the mean, trends in the 95% confidence interval, and distributions of the isotopic densities. Just as in the more familiar Monte Carlo simulations of fluxes or reaction rates, convergence of isotopic densities may exhibit expected behavior such as stable means and 95% confidence intervals that decrease like N{sup -1/2} , or may exhibit poorly-converging behaviors. The most poorly-converging cases in this work were found to be for the products of high-energy threshold reactions. Finally, we examine the conditions under which uncertainty in the isotopic density can be determined from reaction rate uncertainty, and demonstrate the need for considering correlated variates in isotopic uncertainty propagation. Correlation between reaction rate tallies appears most important when the cross sections for the reactions and neutron energies considered are not strongly varying. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
Energy deposition validation results for the evaluated electron data library in FRENSIE - 25357
The Framework for Research in Nuclear Science and Engineering (FRENSIE) is a collection of packages written in C++ that are designed primarily to aid in methods development. A recent development in FRENSIE is the modeling of single scattering electron interactions using the data in the Evaluated Electron Data Library (EEDL) from Lawrence Livermore National Laboratory (LLNL). Preliminary verification and validation (V and V) results using the well-known Lockwood energy deposition experiment and a comparison with MCNP6.2 were performed using several bivariate grid policies (Unit-base, Correlated, Unit-base Correlated), interpolation schemes (lin-lin, log-log, lin-log), and physics options (Coupled and Decoupled Elastic Scattering). The results show an agreement to within 0.5% between MCNP6.2 and FRENSIE when using the same bivariate grid policy (Correlated) and interpolation scheme (log-log). When compared to the experimental results of the Lockwood experiment, it was found that a Unit-base bivariate grid policy, lin-lin interpolation scheme, and lin-log interpolation scheme did not produce accurate results for the energy deposition in test problems. The log-log Correlated and Unit-base Correlated results showed the greatest agreement with the Lockwood experiment. The total energy deposition matched the experimental data to within 3%. Both the Correlated and Unit-base Correlated results had a tendency to estimate a higher energy deposition near the peaks and a lower energy deposition near the tails. The differences between coupled elastic scattering and decoupled scattering models were shown to be more prominent at higher energies with the results of the coupled elastic scattering being more accurate for the peak and less accurate for the tail of the energy deposition profiles. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
Spectral-average cross section validation in LR-0 reactor spectrum of {sup 55}Mn(n,2n), {sup 90}Zr(n,2n) and {sup 127}I(n,2n) reactions - 25362
This article deals with the validation of spectral average cross-sections (SACS) of {sup 55}Mn(n,2n){sup 54}Mn, {sup 90}Zr(n,2n){sup 89}Zr and {sup 127}I(n,2n){sup 126}I reactions in well-defined LR-0 reactor spectrum. The measurement of SACS for selected reactions is of high interest because they are used for the practical reactor dosimetry and also for refinement of prompt fission neutron spectra of {sup 235}U in the region of higher energies. Various experiments for description of neutron field in reactor and detector characterization, which are necessary for validation, are described. The spectral average cross-sections are derived from Net Peak Areas of irradiated samples measured using a semiconductor high purity germanium detector. The spectral average cross section in {sup 235}U prompt fission neutron spectrum was experimentally determined to be 0.239 ± 0.015 mb for {sup 55}Mn, 1.209 ± 0.052 mb for {sup 127}I and 0.107 ± 0.005 mb for {sup 90}Zr. The obtained results were compared with MCNP6 calculations using ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF-1.05 nuclear data libraries. Notable discrepancies were found in {sup 90}Zr(n,2n){sup 89}Zr and {sup 127}I(n,2n){sup 126}I reactions. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Rossi and Feynman-alpha formulas including prompt neutron decay and detector die away time constants - 25396
A simple Rossi-alpha formula and the corresponding Feynman-alpha formula have been derived that take into account two time constants; one characterizing the prompt neutron decay in the fissile source domain and the other the detector die-away time. The derivation of the formula is an extension of the classic heuristic method. A test problem is defined and simulated using the Monte Carlo MCNPX code in order to illustrate the applicability of the formulas. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
Comparison of MAVRIC/MONACO and TRIPOLI-4{sup R} simulations of the LLNL pulsed spheres benchmark experiments - 25403
Simulations of the Lawrence Livermore Pulsed Spheres with MAVRIC/Monaco and TRIPOLI-4{sup R} are presented. Both codes use ENDF/B-VII.1 cross section data, but TRIPOLI-4{sup R} also uses cross section data based on JEFF-3.1.1. The MAVRIC/Monaco simulations were performed with the SCALE 6.2.2 release while the TRIPOLI-4{sup R} simulations used an internal development version of the code. For the most part, the simulated and measured neutron counts agree within 30%. Simulated gamma flux data is compared, as well. There are no measured gamma data. The agreement between the simulated gamma flux spectra is often within 10% or less, but there some notable differences, especially when comparing simulations using ENDF/B-VII.1 and JEFF 3.1.1 based cross section data. The comparison of photon simulations suggests that improvements are needed in the cross-section data, cross section processing codes, and/or transport codes and gamma measurement data is needed for experimental validation. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
First assessment of the new atomic data used in ENDF/B-VIII - 25420
This conference paper summarizes an initial assessment of the evaluated atomic data libraries released in ENDF/BVIII. 0 and in EPICS2017 in early 2018. The new versions supersede the data libraries currently used by Monte Carlo particle transport codes to handle electron and photon interactions with matter, which date back to the 1990's. Several features of the data are evaluated: their consistency, the content available in the different formats of the data libraries and from different distribution sources, and the computational performance associated with their use. Additionally, an overview of the validation status of the data is given, including some preliminary results concerning new cross section data released in 2018. These evaluations contribute to estimate the impact of the new data libraries in Monte Carlo simulations and highlight areas where future versions would benefit from improvements. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
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↗
The decay data sub-library in ENDF/B-VIII.0 - 25453
We describe the ENDF/B-VIII.0 decay data sub-library, in particular the main changes since its previous release. Its application in calculating the antineutrino spectra from nuclear reactions is briefly presented. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗
Specializations in the SCEPTRE code for charged-particle transport - 25454
Charged particles present some unique challenges for radiation transport codes. This is because charged particles have cross sections that are extremely forward peaked, are huge in the limit of small energy transfer, and are highly scattering, which causes slow convergence of the source iterations. The primary application of SCEPTRE is modeling radiation-driven electrical effects, so substantial effort has been invested in SCEPTRE for the efficient modeling of electron transport. This paper will summarize recent and ongoing activities involving the accurate deterministic-transport modeling of charged particles and methods implemented to improve iterative convergence. (authors)
61 RADIATION PROTECTION AND DOSIMETRY↗