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At least 1,801 records · Page 100
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Increasing Interlaminar Strength in Large Scale Additive Manufacturing

Interlaminar strength of extrusion-based additively manufactured parts is known to be weaker than the strength seen in the printed directions (X and Y). With Big Area Additive Manufacturing (BAAM), large parts lead to long layer times that are prone to splitting, sometimes referred to as delamination, between the layers. Fiber filled materials, such as carbon fiber reinforced ABS, are used to counter-act the effects of thermal expansion, but the fibers stay in-plane meaning that no fibers span from layer to layer, which helps counteract the weak interlaminar strength that causes splitting. A solution to this is a patent pending approach known as “z-pinning”. The process involves strategically positioning voids across multiple layers that will be backfilled with hot extrudate. This paper will explore the benefits and results of using “z-pinning” in large scale additive manufacturing.

Roschli, Alex↗
DOE OSTI2017

Reducing mechanical anisotropy in extrusion-based printed parts

The mechanical performance of 3D printed components is highly dependent upon the orientation of the part relative to the build plane. Specifically for extrusion-based printing systems, the bond between successive layers (z-direction) can be 10-25% weaker than in the printed plane (x-y plane). As advanced applications call for fiber reinforced materials and larger print systems (such as the Big Area Additive Manufacturing system) extend the layer time, mechanical performance in the z-direction can decrease by 75-90%. This paper presents a patent-pending approach for improving mechanical performance in the z-direction by depositing material vertically across multiple layers during the build. The “z-pinning” process involves aligning voids across multiple (n) layers, which are then back-filled in a continuous fashion during the deposition of layer (n+1). The “z-pinning” approach has been demonstrated to be an effective approach for increasing the strength (20% increase) and toughness (200% increase) of printed parts in the zdirection.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Modeling of the TREAT M8 Calibration Series Measurements using MAMMOTH

This presentation provides an overview of the modeling efforts performed in to validate the MAMMOTH multi-physics reactor analysis application using historical TREAT data. These data come from the M8 series of calibration measurements, which was the last series performed before TREAT was shut down in 1994. The data applied in this work include an experiment flux profile and low level steady state power coupling factors for both flux wires and fuel pins. MAMMOTH was able to obtain reasonable results for all cases with the exception of the H91-8-1 flux wire, the only flux wire containing highly-enriched uranium. The feasibility of performing similar simulations of the calibration series fuel pins was demonstrated with the flux wire calculations, and will be performed in future work.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Capacitance-Based Dimensional Change Sensors for In-Pile Materials Measurements

Advanced fuel compositions, such as accident tolerate fuels (ATF), are an active area of developing in the nuclear power industry. The long-term performance of these newly developed fuels is estimated through physics-based simulation models of irradiation-, temperature-, pressure-, etc.-induced material degradation. As these fuels are deployed in test reactors, measurement and characterization of the fuel pin evolution is used to validate prediction models. In-pile material evolution parameters, such as fuel rod pressurization, fuel stack and cladding elongation, and cladding diameter, are commonly measured using a linear voltage differential transformer (LVDT). However, LVDTs are bulky and limited to lower (350-500C) temperature operation. The high power density and small size of most experimental positions in high performance research reactors used for accelerated materials irradiation studies generally precludes the use of LVDTs in these reactors. There is a critical need for sensors that provide real-time data regarding material evolution under highly accelerated irradiation. These sensors would ideally have a small profile and the ability to withstand irradiation at extremely high dose rates and temperatures for extended periods of time. A capacitance-based sensor is currently under development at the University of Tennessee to provide a direct measurement of in-pile dimensional change during irradiation. Sensor response was simulated using AutoCAD Electromagnetic field simulator (EMS) for a variety of sensor materials and configurations and fuel pin swelling conditions. Initial results of these simulations are summarized and areas of ongoing research and development are discussed.

Naughton, Tyler↗
DOE OSTI2013

Zener diode protection network in submount for LEDs connected in series

A transient voltage suppressor circuit is disclosed for a plurality (N) of LEDs connected in series. Only one zener diode is created for connection to each node between LEDs, and a pair of zener diodes (the “end” zener diodes) are connected to the two pins (anode and cathode pads) of the series string. Therefore, only N+1 zener diodes are used. The end zener diodes (Q1 and Qn+1) effectively create back-to-back zener diodes across the two pins since the zener diodes share a common p+ substrate. The n+ regions of the end zener diodes Q1 and Qn+1 have the highest breakdown voltage requirement and must be placed relatively far apart. Adjacent n+ regions of the intermediate zener diodes have a much lower breakdown voltage requirement so may be located close together. The zener diodes may be placed within a very small footprint or can be larger for better suppressor performance.

Wei, Yajun↗
DOE OSTI2019

Implementation of Gamma Transport Capability in MPACT

This paper presents the latest work on the development of the gamma transport capability for the CASL neutronics code MPACT. MPACT is capable of performing pin-resolved neutronics calculations for full-core problems and the purpose of this project is to further expand its capability for gamma transport calculations at the same level of accuracy for potential applications of interests, such as gamma heating and gamma fluence calculations. The development of the gamma transport capability reuses many existing geometrical capabilities and neutronics solvers in MPACT, because essentially the same form of Boltzmann transport equation is solved. First, 2-D gamma transport is implemented in this work by leveraging the method of characteristics (MOC) sweeper. The HELIOS library is used to provide the neutron-induced gamma production and gamma interaction cross sections. The new gamma transport capability is verified with MCNP. Both the gamma spectrum of a pin cell problem and the gamma flux distribution of 2D assembly problems are consistent with reference calculations. A 2D quarter core case is qualitatively verified by comparing MCNP and MPACT predictions of the gamma flux and thermal neutron flux to confirm the gamma smearing effects.

Wang, Xinyan↗
DOE OSTI2018

Initial Investigation of the Small Feature Detection Capability of the ORNL Fast Neutron Tomographic Imaging System

A variety of fast neutron imaging simulations and measurements were performed to investigate the spatial resolution and small feature detection capability of a fast-neutron tomographic imaging system developed at Oak Ridge National Laboratory. The system uses an associated particle DT (deuterium-tritium) neutron generator as the source and 25 mm × 25 mm × 100 mm plastic scintillators as the neutron detectors. Two image quality indicators were examined: 1) a variety of long steel pins embedded axially in the interior of a polyethylene cylinder and 2) a variety of short steels pins embedded radially in the surface of a shielded polyethylene cylinder. Measurements demonstrated detectability down to approximately 3 mm. Further data analysis may show improved detection performance and allow a more precise determination of the spatial resolution and small feature detection limit. The current DT generator has a neutron production spot size of about 3.5 mm, which appears to be the limiting factor in the current system configuration.

Wright, Michael C.↗
DOE OSTI2020

Nuclear reactor support and seismic restraint with core retention cooling features

A nuclear island includes a nuclear reactor, a lateral seismic restraint, and a reactor core retention cooling system. The lateral seismic restraint includes a vertically oriented pin attached to one of the bottom of the lower vessel head and the floor underneath the nuclear reactor, and a mating pin socket is attached to the other of the bottom of the lower vessel head and the floor. The reactor core retention cooling system includes one or more baffles, optionally thermally insulating material, disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head. A plenum is defined between the one or more baffles and the exterior surface of a lower portion of the reactor pressure vessel. The one or more baffles may define a lower plenum inlet surrounding the lateral seismic restraint.

Edwards, Tyler A.↗
DOE OSTI2019

The Fission Accelerated Steady State Test (FAST) -- A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

In an effort to improve the development and testing of novel nuclear fuels, a detailed evaluation on reducing the size of fuel specimens for irradiation has been performed. Theoretically speaking, reducing the fuels diameter and incorporating an inner capsule while maintaining the original outside diameter of the experiment significantly reduces the experiment’s sensitivity to fabrication tolerances and drastically reduces the time necessary to reach a desired fuel burnup; both of which improve the cost and schedule associated with the fuel testing. Simulations of this reduced diameter, double encapsulated concept have been performed using Abaqus, MCNP, and ORIGEN. The results of the Abaqus simulations are beyond the scope of this summary and will not be presented herein. For results of these Abaqus simulations, see Beausoleil [2]. The models were tested using various parameters and design deviations which were then compared to the original Advanced Fuels Campaign (AFC) designs. The results of these simulations confirmed that the introduction of a second capsule coupled to a reduction in the fuel pin diameter significantly reduces the experiment’s sensitivity to fabrication tolerances and reduces the time to reach a desired burnup by nearly an order of magnitude. The original AFC design requires approximately 11.7 years to reach 30 Atom % Heavy Metal (at% HM) burnup. A burnup of 30 at% HM can be reached in approximately 2.5 to 3.5 years (irradiation position dependent) with a fuel diameter ½ (~0.097 in. diameter) that of the original AFC design. An added benefit to the fuel pin diameter reduction is that more specimens can be irradiated in any given cycle. Essentially, significantly more specimens can be tested and incur more burnup over a given operating cycle relative to the AFC design. This allows the development of a large test matrix of samples with rapid throughput in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

EVALUATION OF PRESSURE DROP CORRELATIONS FOR THE WIRE-WRAPPED ROD BUNDLES

The accurate prediction of pressure drop in the fuel rod bundle is of importance for the design and thermal-hydraulic analysis of nuclear reactor system. Due to very complex geometry of wire-wrapped fuel rod bundle geometry for the Sodium Fast Reactor (SFR), a simple method using hydraulic (equivalent) diameter is not sufficient to predict the pressure drop of rod bundle accurately. For several decades, many experimental works have been performed to develop an accurate correlation for wire-wrapped fuel bundle friction factor. In these works, the friction factor correlation of wire-wrapped fuel bundle has been developed as a function of Reynolds number of the flow and the design parameters such as fuel pin diameter, wire diameter, rod pitch, and wire lead length. The Cheng and Todreas correlations are well-known and used in many subchannel analysis code, such as CADET, TRIO-U, MATRA-LMW, SE2-ANL, etc. Novendstern model, Rehme model, Baxi and Dalle Donne model are also well-known and widely referred. Computational Fluid Dynamics is another typical approach to predict the pressure drop of wire-wrapped fuel rod bundle, but it requires the model validation because the accuracy of pressure drop prediction by the CFD simulation depends on what turbulence model is employed. The present work aims at evaluating the existing pressure drop correlations for the wire-wrapped fuel rod bundle and evaluating the pressure drop of 217-pin rod bundle by the CFD analysis.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

PRODUCTION OF GLASS-TO-METAL SEALS USING ADDITIVE MANUFACTURING TECHNIQUES

Glass-to-metal (GtM) seals are hermetic barriers between glass and metal components, often used in the electronic and vacuum industries. Creating the seals traditionally requires heating all the components to the glass melting temperature, where the glass will flow and bond to the metal. Due to the manufacturing constraints of the individual components and sealing conditions there are geometric and material restrictions. Additive manufacturing techniques were used to make GtM seals to reduce these restrictions. Hermetic, single pin seals were produced using printed metal shells, by selective laser melting (SLM), and digital light processing (DLP) printed glass preforms. Glass preforms were printed with photosensitive resin mixed with a silicate sealing glass powder in a 1:2 weight ratio. Thermal analysis, TGA/DSC, and screening experiments were used to determine an organic removal schedule. Video techniques determined a sintering schedule by producing a densification curve at different isothermal holds which were fit with a viscous sintering model. Single pin seals were also produced use a laser assisted manufacturing (LAM) process as an alternative to the conventional furnace sealing process. A Nd:YAG laser was focused on a solid glass preform, heating the glass to form the glass-metal bond. The sealing glass was doped with a transition metal oxide, Fe2O3 or CuO, to increase absorption of the laser wavelength. The interface of hermetic LAM seals were analyzed using electron microscopy, compared to conventional prepared seals, finding a reduction of interfacial morphologies.

Read, Aaron Reece↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Self-shielding models of MICROX-2 code

The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. In the previous study, a new fine-group cross section library of the MICROX-2 was generated and tested against reference calculations and measurement data. In this study, existing physics models of the MICROX-2 are reviewed and updated to improve the physics calculation performance of the MICROX-2 code, including the resonance self-shielding model and spatial self-shielding factor. The updated self-shielding models have been verified through a series of benchmark calculations against the Monte Carlo code, using homogeneous and pin cell models selected for this study. The results have shown that the updates of the self-shielding factor calculation model are correct and improve the physics calculation accuracy even though the magnitude of error reduction is relatively small. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by approximately 0.1 % and 0.2% for the homogeneous and pin cell models, respectively, considered in this study. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Evaluation of the use of color-set geometry during lattice physics constants generation for boiling water reactor simulation

Current methods for BWR nuclear design and analysis consist of using lattice physics neutron transport methods to generate the two-group homogenized cross-sections that are then used in a nodal diffusion theory code. The lattice transport solutions are performed for a single assembly with reflective boundary conditions, which is a practical approximation. A method is developed to account for assembly exposure distributions (environment) in the core within the lattice transport calculations with the use of color-sets (2x2) geometry. The loading pattern is examined and an appropriate number of characteristic color-set cells are selected for analysis. Treatment of the co-resident exposed fuel within this method is also presented. The calculation process was followed for a recent BWR cycle design with comparisons being performed on both a lattice and core-wide basis to evaluate the proposed method. The lattice based comparisons show noticeable differences in the pin power distribution predictions, which require further investigation to see how this translates into core performance calculations. The core-wide comparisons show minor differences and are generally in a good agreement, which is expected with this small perturbation. A slight improvement was noticed in the reduction of the power distribution uncertainty. However, given the additional amount of work and computer run time increase, further evaluation, especially of core pin power predictions, is needed to consider this method for production level design and safety analysis calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Improvements, enhancements, and optimizations of COBRA-TF

The Reactor Dynamics and Fuel Management Group (RDFMG) at The Pennsylvania State University (PSU) has become active in the Consortium for Advanced Simulation of Light Water Reactors (CASL) program by delivering, supporting, and further developing CTF, the PSU version of the Coolant Boiling in Rod Arrays - Two Fluids (COBRA-TF) Thermal/Hydraulic (T/H), sub-channel program. New development work on CTF was primarily geared towards decreasing the execution time of the code so that it may eventually be used for performing pin-by-pin, full-core simulations. Great gains have been made through targeting sections of source code for optimization. For example, wall clock time has been reduced for a one-assembly, three-dimensional model, containing {approx}9,400 mesh cells, from 9.2 min to 1 min. A further improvement has been reduction in code memory usage, which was previously prohibitive for large models. In conjunction with the run time speedups, this has enabled the simulation of a refined quarter-core model ({approx}460,000 mesh cells), which saw a reduction in memory usage from over 130 GB to less than 3 GB. In addition to the optimization work, RDFMG has also created a preprocessor utility for the fast and less error-prone generation of CTF input decks. Furthermore, basic post-processing capabilities have been implemented by creating a CTF subroutine for producing Visualization Tool-Kit (VTK) files that output mesh data and associated simulation results. These VTK files can be opened with visualization software. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

A demonstration of a whole core neutron transport method in a gas cooled reactor

This paper illustrates a capability of the whole core transport method COMET. Building on previous works which demonstrated the accuracy of the method, this work serves to emphasize the robust capability of the method while also accentuating its efficiency. A set of core configurations is presented based on an operating gas-cooled thermal reactor, Japan's HTTR, and COMET determines the eigenvalue and fission density profile throughout each core configuration. Results for core multiplication factors are compared to MCNP for accuracy and also to compare runtimes. In all cases, the values given by COMET differ by those given by MCNP by less than the uncertainty inherent in the stochastic solution procedure, however, COMET requires runtimes shorter on the order of a few hundred. Figures are provided illustrating the whole core fission density profile, with segments of pins explicitly modeled individually, so that pin-level neutron flux behavior can be seen without any approximation due to simplification strategies such as homogenization. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

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