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At least 127 records · Page 7

Development and Evaluation of Neutron Thermalization Integral Benchmarks for Advanced Reactor Applications (Final Report)

In this project, integral benchmarks that aim to examine thermal neutron scattering data for graphite (ideal and nuclear), light water, and molten salt FLiBe will be analyzed. In particular, the project aims to: 1) Develop the slowing-down-time experiment that was performed at the OREAL facility of ORNL as a graphite benchmarks; 2) Identify and examine a slowing-down-time experiment in light water as a potential benchmark; 3) Reexamine the molten salt MSRE-MSR-EXP-001 benchmark to explicitly capture the impact of TSL variations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of Transmission Measurements Capability at VISION Spectrometer [Slides]

This presentation covers the development of transmission measurements capability at VISION Spectrometer. Main takeaways include that INS and transmission measurements are and should be the most significant part of TSL validation, as they are the fundamental quantities used by Monte Carlo neutronics codes. Additionally, the development of transmission capability at VISION instrument has commenced. Promising first results have been obtained with a big impact by the use of the collimators. Finally, Areas of improvement have been identified and new measurements are planned for June.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Evaluation of Thermal Neutron Scattering Cross Section of Uranium Silicide with Ab Initio Lattice Dynamics

Uranium silicide (U 3 Si 2 ) is a candidate material for the high-density nuclear fuel in commercial light water reactors [1], [2]. Its higher uranium density, 11.3 g-U/cm3, compared to that of uranium dioxide (UO 2 ), 9.7 g-U/cm3, can improve the performance of a nuclear reactor while using low enriched uranium (LEU) and diversify the choice of cladding materials [1]–[3]. It also has a higher thermal conductivity than UO 2 , which can reduce the thermal stress on the material caused by a temperature gradient across the fuel pellet and provide a larger margin for some postulated accidents [1], [2], [4]. Furthermore, compared to U3Si, another high-density fuel candidate, it has better resistance to in-pile swelling due to less irradiation-induced rapid amorphization [1], [3]. Corresponding to its importance in nuclear engineering, many previous studies have reported the properties of U3Si2. Experiments showed that U 3 Si 2 is a paramagnetic (PM) metal, where a slight linear increase in magnetic susceptibility was measured with increasing temperature [5], [6]. In addition, thermodynamic quantities such as thermal expansion coefficient, heat capacity, and thermal conductivity were experimentally determined over a wide temperature range [1], [7], [8]. In several computational studies, ab initio atomistic simulations based on density functional theory (DFT) were performed to calculate various properties including elastic constants, electronic density of states (DOS), and phonon dispersion curves [9]–[12]. Nevertheless, thermal neutron scattering cross sections, which are critical to the prediction of the parameters in reactor physics that are ultimately related to reactor criticality, have not yet been evaluated for U3Si2. The scattering cross section can be calculated from the phonon DOS, or the energy spectrum of lattice vibrations, of the crystalline system [13], [14]. However, there is also no experimental data available for the phonon DOS of U 3 Si 2 . While some computational studies reported the phonon DOS and/or dispersion curves from ab initio simulations [9]–[12], the accuracy cannot be guaranteed because it is unclear whether the spin-polarization behavior of PM U 3 Si 2 was properly described. In the present study, the thermal neutron scattering cross section for U 3 Si 2 is evaluated for the first time by calculating the phonon DOS for U3Si2 from ab initio lattice dynamics (AILD) simulations based on DFT. First, U 3 Si 2 is modeled based on the experimental structure, and AILD simulations are performed on the modeled U3Si2 to optimize the structure. Next, AILD simulations are performed for supercells with atomic displacement to calculate Hellmann-Feynman forces. Based on the calculated forces, partial phonon DOSs for U and Si are obtained, and the thermal neutron scattering law (TSL) for U 3 Si 2 is finally evaluated. To verify the accuracy of the calculations in the present study, the calculation results are compared with experimental data on the structure and heat capacity of U3Si2 [1], [7], [8], [15].

Geometry Optimization↗

Integral Experiment Request 523 CED-2 Report

This report documents the final design phase of the Critical Experiment Design (CED-2) conducted as part of integral experiment request (IER) 523. The purpose of IER 523 is to determine critical configurations of 35 weight percent (wt%) enriched uranium dioxide beryllium oxide (UO 2 -BeO) material driven by an annular ring of Seven Percent Critical Experiment (7uPCX) fuel rods at Sandia National Laboratories (Sandia). The experiments will provide benchmark data on water moderated, intermediately enriched UO2 systems as well as Be nuclear data. The experiment will also provide partial validation for the beryllium oxide (BeO) thermal neutron scattering law (TSL) in the thermal energy range. Experiment design concepts, neutronic analysis results, and proposed paths for continuing the CED process are presented. This report builds on the feasibility and justification of experimental need report (CED-0) and preliminary experiment design report (CED-1) completed in December 2021 and September 2023, respectively [1, 2].

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Design of Critical Experiments Involving Graphite elements at the IPEN/MB-01 Plate-Type Core

A series of critical experiments were designed involving the use of characterized graphite elements in the new plate-type core of the IPEN/MB-01 reactor. Eleven configurations with different graphite quantities are considered, each producing distinct effects on neutron fluxes and offering opportunities to analyze varying sensitivities to k eff in the system. The analysis shows that the experiments can be performed with acceptably low k eff uncertainties considering an adequate characterization of the graphite elements to be used. The sensitivity results obtained using SCALE TSUNAMI calculations show that the experiments have a notable k eff sensitivity coefficient to graphite. In the configurations tested, the k eff sensitivity to graphite TSL is low. This experiment will be useful for criticality safety validation of applications using uranium fuel enriched to around 19.75 wt % 235 U, light water and graphite. It will also be useful to gain more insight into graphite material properties and their effect on k eff as a result of using the well-characterized graphite elements. The execution of the experiments is planned for the summer of 2025. The 11 designed configurations serve as a basis for the final design, and fewer configurations will be executed. Once executed, these critical experiments will be evaluated and submitted for publication in the International Criticality Safety Benchmark Evaluation Project Handbook, supporting the DOE/NRC Collaboration for Criticality Safety Support for Commercial-Scale HALEU for Fuel Cycles and Transportation project goal.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Continuing Development of the Nuclear Data Processing Code AMPX [Poster]

The ENDF/B-VIII.1 evaluation library has seen a great growth in the thermal neutron scattering sub-library. The SCALE code system has traditionally approached CE transport by assuming that the CE library on disk represented the fully expanded cumulative probability distributions, conditional on exiting angle and marginal on exiting energy. While this is a complete description of the data, it comes at the potential cost of large amounts of on-disk storage. This approach was strained by several TSL files in ENDF/B-VIII.1, such as graphite, which contained data for a large number of Bragg edges. In the fully expanded probability distributions, this was found to be a disproportionately large fraction of the SCALE CE library.

GNDS↗

Trajectory Specification Language for Air Traffic Control

Trajectory Specification is a method of specifying aircraft trajectories with tolerances such that the position at any instant in time is constrained to a precisely defined bounding space. The bounding space at any instant in time is defined by tolerances relative to a reference trajectory that specifies position as a function of time. The tolerances are dynamic and are based on the aircraft navigation capabilities and the traffic situation. This paper proposes a standard Trajectory Specification Language (TSL) based on the Extensible Markup Language (XML) to represent these specifications and to communicate them by datalink. The language can be used to downlink trajectory requests from air to ground and to uplink trajectory assignments from ground to air. The XML format can be converted to binary for operational use, if necessary, using Efficient XML Interchange (EXI) or Abstract Syntax Notation (ASN.1).

XM↗

Near-Saturation Conditions at the Tropical Tropopause: Results from Ticosonde

The TTL lies between the neutral buoyancy level (NBL) at ~350 K and the tropopause. Within the the TTL radiative heating drives ascent and air parcel supersaturation. Ticosonde measurements since 2005 show that the incidence ofsupersaturation in the TTL over Costa Rica is ~60%. This is due to diabatic ascent. The frequency is highest (68%) in summer, when convection is frequent. The TSL was defined by Selkirk et al. (2010) as the upper edge of the TTL. It isIn this layer that the final saturation of air parcels rising into the stratosphere occurs and thus the water vapor minima which define the so-called "writehead" of the Atmospheric Tape Recorder.

Selkirk, Henry↗

Development of a Combined Cohesive and Virtual Crack-Closure Technique Approach to Represent R-Curves

Resistance curve (R-curve) effects due to fiber bridging, crack migration, and other blunting mechanisms are common in composite laminates. These mechanisms can dramatically increase the resistance to crack propagation but can be challenging to simulate. Delamination with R-curves can be analyzed using cohesive zone modeling (CZM) or the virtual crack closure technique (VCCT). Large fracture process zones can be simulated with CZM, but they require highly refined meshes. Coarser meshes can be used with VCCT, but this method is only applicable to small fracture process zones with R-curves defined as functions of position. Therefore, a technique with the computational efficiency of VCCT and the natural ability of cohesive elements to represent large fracture process zones is desirable. An approach is proposed that starts as a CZM in which the cohesive traction separation law (TSL) is separated into high-strength (HS) and low-strength (LS) components. The HS part, which is responsible for the mesh requirements of the CZM analysis, is replaced by VCCT. The combined method is evaluated by analyzing the response of double cantilevered beam specimens. Two specimen layup configurations, [0/90/90/0] 3s and [0/90/90/0] 9s , are evaluated. The R-curve response from each configuration is determined and applied to the constitutive properties of models built with CZM and with VCCT. The results and computational efficiency of the CZM, VCCT, and combined approaches are compared. The results are indicative that a combined cohesive/VCCT approach can enable progressive failure analyses to retain the computational efficiency of VCCT with the ability of the cohesive elements to capture R-curve effects.

Virtual Crack-Closure Technique↗

Next-generation system analysis model for studying alternative Waste Management Systems - 15232

The U.S. Department of Energy's Office of Nuclear Energy (DOE-NE) is conducting planning activities within the Nuclear Fuels Storage and Transportation Planning Project (NFST). The plans involve (1) laying the groundwork for implementing interim storage (including dealing with any associated transportation that is needed) per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste of January 2013 and (2) developing a foundation for a new nuclear waste management organization. An overarching goal is to develop options for designing an integrated system that will be considered by decision-makers. Based on the dynamic system analysis and software specification requirements that must be met, Argonne National Laboratory (Argonne) is leading the development of a consolidated waste management system tool known as the Next-Generation System Analysis Model (NGSAM). The development of NGSAM leverages the approaches and methods used previously to develop the Transportation Storage Logistics (TSL) and Total System Model (TSM) system analysis tools, but the new tool takes advantage of advanced computational capabilities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

U.S. Department of Energy, Office of Nuclear Energy, Nuclear Fuels Storage and Transportation Planning Project Modeling Tools

The U.S. Department Energy, Office of Nuclear Energy's (DOE-NE) Nuclear Fuels Storage and Transportation (NFST) is developing foundational capabilities and tools to support and inform the future deployment of consolidated interim storage and large-scale transportation of SNF. These tools include Used Nuclear Fuel Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-Standards), Stakeholder Tool for Assessing Radioactive Transportation (START), Transportation and Storage Logistics (TSL), Next- Generation System Analysis Model (NGSAM), Execution Strategy Analysis (ESA), Multi-Objective Evaluation Framework (MOEF), and the Centralized Used Fuel Resource for Information Exchange (CURIE). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Ab initio generation of thermal neutron scattering law for uranium dioxide

Neutronic analysis of moderated nuclear fission systems, e.g. nuclear reactors, relies upon an accurate prediction of the thermal neutron spectrum. Thermal neutrons have wavelengths on the order of the interatomic spacing between atoms in matter. Consequently, the scattering cross-sections of these neutrons are affected by the scattering law of the material (i.e. S(α,β), where α and β are dimensionless momentum and energy transfer variables), which describes the atomic structure and thermodynamic motion of the atomic nuclei. Uranium-dioxide (UO{sub 2}) is the primary fuel source for nuclear power reactors moderated by light water. In such systems (e.g. nuclear power reactors, spent fuel pools, etc.) the nuclear fuel comprises a large fraction of the system volume, approximately 1/3 in light-water reactors. Therefore, the scattering law of the fuel material may impact the thermal neutron spectrum and criticality of these systems. In this work previously established ab initio lattice dynamics (AILD) methods were utilized to calculate the phonon density of states (DOS), which acts as a fundamental input to the scattering law. Subsequently, the S(α,β) of UO{sub 2} was generated for publication as an ENDF thermal scattering library (TSL), which will be available for use in standard neutronics computer applications. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Investigation of Treat M2 and M3 calibration experiments for benchmark analysis

The Transient Reactor Test Facility (TREAT) located at Idaho National Laboratory is a graphite moderated and reflected, air cooled core. This reactor was originally designed and operated to experimentally test fuel under extreme transient conditions. To this end, TREAT successfully performed many series of steady-state and transient tests until operations were discontinued in 1994. With the restart of the reactor planned for 2018, accurate models of previous reactor experiments are necessary to verify the ability to predict conditions for planned upcoming tests. In this work, the methodology for modeling the M2 and M3 calibration experiments (M2/M3CAL) will be described along with results from the variation of key parameters including the boron concentration. Additionally, this model will be used to demonstrate the correlation between reactor behavior and the implementation of 'nuclear graphite' thermal scattering law (TSL) data to capture the behavior of thermal neutrons in the TREAT core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Luminescence dosimetry for evaluation of the external exposure in Metlino, upper Techa River valley, Southern Urals, Russia: Analysis of new results

Luminescence dosimetry was applied in the former settlement of Metlino, Southern Urals, Russia as part of a full-scale study to validate the Techa River Dosimetry System (TRDS) 2016 for the upper Techa River region. The village, which was evacuated in 1956, was located 7 km downstream of the release point of liquid radioactive waste by the Mayak plutonium facility. Several brick samples were taken from north-eastern and south-eastern walls of the granary, facing the former Techa river shoreline and floodplain. Samples were all taken at the same height and measured at different depths into the brick. For the majority of brick samples, good Optically Stimulated Luminescence properties of the quartz grains were observed. In some cases, however, strong levels of sensitization and/or signal recuperation were encountered which necessitated adjustment in the measurement protocols. Anthropogenic doses in bricks varied from 1.5 to 6.6 Gy and the horizontal profiles along both walls showed significant variation, which is explained on a qualitative basis. Further, a dose depth profile is observed for selected samples, which is different from the dose depth profile measured and simulated for samples from the north-western wall of the granary in previous studies. This is qualitatively explained by the differences in source configuration.

61 RADIATION PROTECTION AND DOSIMETRY↗

Preliminary Design of a New Pulsed-Neutron Die-Away Experimental Testbed for Thermal Scattering Law Benchmarks (PNDA) (IER-501 CED-1)

This report presents the preliminary design (CED-1) of IER-501, a new testbed for pulsed-neutron die-away (PNDA) experiments at Lawrence Livermore National Laboratory (LLNL). The PNDA experiments will be high-quality benchmarks that serve to optimize and validate thermal scattering laws (TSLs), which are high-priority nuclear data for the Department of Energy’s Criticality Safety Program (NCSP). PNDA experiments can be used to generate nuclear data validation benchmarks quickly and cheaply, as they do not require fissile material or a nuclear facility to conduct them.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗