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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Benchmark Exercise for the Control Rod Swelling Evaluation

The VTR core has six reactivity control assemblies and three safety assemblies. The control assemblies or primary control rods are adjusted during the normal operation to balance the core reactivity and to control the reactor power. A typical control assembly radial layout is presented in Figure 1. The figure shows the swelled absorber (B 4 C) rod. Initially, helium gas fills the gap between the pin and the cladding before irradiation swelling takes place. For VTR, HT9 steel was selected as the cladding and duct material. The main neutron absorbing material used in the VTR is B 4 C. When residing in the core, the neutronics, thermophysical, and mechanical properties of the materials used in a control assembly will degrade due to accumulated neutron damage. Material degradation limits how long a control assembly can reside in the core. Many phenomena affect the control assembly lifetime, such as the loss of reactivity worth due to B 4 C depletion, the mechanical interaction of the absorber rod and the cladding due to B 4 C swelling, the helium gas buildup in the pin due to B-10 capture, etc. B 4 C swelling, which causes closure of the gap between the absorber rod and the cladding, is usually considered as the main limiting factor from past experience. An initial study was conducted at PNNL to evaluate the irradiation behavior of a VTR control assembly. The evaluation was performed using the CNRD2 code that was initially developed for the FFTF. The study also included an assessment of the VTR control assembly and focused on a 61-pin control assembly design, which is different from that used (37-pin design) in the core design study. The study conducted by PNNL was reviewed independently by ANL. A Python script referred to as the Control Assembly Evaluation Script (CAES) was developed for the independent review and additional assessment of 37-pin control assembly design. The script has focused on the assessment of the absorber rod swelling for its importance in determining the control assembly lifetime. CAES uses geometry, neutronics, materials data as input to predict the swelling of the absorber rod during its residence in the reactor core. The results from CAES showed some non-negligible differences against the PNNL results. Some of the differences can be attributed to the different interpretation of the control rod assembly dimensions. To resolve this issue, a benchmark exercise was proposed. The benchmark specification was developed by PNNL. The benchmark exercise was performed independently at PNNL and ANL using different codes/scripts (CRND2 and CAES). This memo documents the results calculated using the different codes. However, this report is limited to presenting the results obtained. Further investigation of the cause of the observed difference will be performed as part of future activities, pending continuation of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

MOOSE Reactor Module Meshing Enhancements to Support Reactor Physics Analysis

The U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) program develops an integrated suite of advanced reactor physics tools built upon the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Each code generally requires an input finite element mesh on which the physics solution is calculated, reported, and transferred to other physics codes. The meshing process is often burdensome for the complex geometries present in reactors due to lack of easy-to-use, open-source meshing tools. To address the bottleneck associated with meshing complex geometries found in nuclear reactors, NEAMS initiated the development of the MOOSE Reactor Module starting in FY21. The Reactor Module builds off the existing MOOSE Mesh System to include targeted meshing capabilities such as the ability to generate hexagonal pin cells, assemblies with ducts, rotating control drums, cores, peripheral zones around a core, as well as the automatic labeling (“reporting IDs”) of pin, assembly, and planar zones to simplify post-processing of results. As a Physics Module in MOOSE, the Reactor Module is open-source, available with any MOOSE installation, directly compatible with MOOSE-based tools, and can be invoked from MOOSEbased applications to generate meshes. Functionality from the Reactor Module has been applied to several advanced reactor concepts to demonstrate user workflow improvements and accuracy. The primary objective of the Reactor Module is to improve useability of MOOSEbased tools by streamlining mesh generation and output inspection processes. During FY22, the functionality of the Reactor Module (and accompanying Mesh System) has been expanded based on user needs. First, the Reactor Geometry Mesh Builder capability developed primarily in FY21 has been refactored and merged to the public MOOSE repository. This capability wraps underlying Reactor Module mesh generators into a “Pin – Assembly – Core” workflow appropriate for conventional Cartesian and hexagonal geometries, and notably assigns material IDs during mesh generation stage and generates only the minimal number of blocks needed in order to reduce computational burden. Biasing and boundary layer options have been added to the base mesh generators as required by thermal hydraulics solvers. The reporting ID functionality has been expanded to differentiate ring-wise and azimuthal sectors within a pin for use with depletion algorithms, and VectorPostProcessor and Reporter objects are now available to integrate solution variables across zones based on ID combinations. Functionality to trim hexagonal meshes along the center or periphery has been developed so users may leverage symmetry and reflective boundary conditions to reduce the mesh size. A flexible and powerful tool to fill the space between two sidesets has been introduced to the framework and can be used for transition layers such as stitching two assemblies together with different numbers of pins, or for complex geometries which do not follow conventional Cartesian/hexagonal patterns. Finally, additional verification problems were performed with NEAMS physics tools in complement with existing NEAMS work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Modeling Flow and Particle Deposition in a Spent Nuclear Fuel Assembly

CFD (Computational Fluid Dynamic) simulation of aerosol-laden natural convective flow and particle deposition in a spent fuel storage canister with 37 assemblies is currently computationally prohibitive. PWR (Pressurized Water Reactor) assemblies have up to 289 pins or tubes with several spacer grids to align the pins. Spacer grids with mixing vanes induce swirling during operation to increase heat transfer. Each spacer grid contains hundreds of small structures such as retaining clips, channel walls, and openings. The largest canisters store 37 PWR assemblies thus, there are numerous pins, tubes, and spacer grids for which the flow region between and around these structures needs to be determined along with the movement and deposition of aerosol particles. Because of the complicated geometry, modeling the intricate flow even for just one assembly is currently impractical. Nonetheless, we are developing techniques for a practical model to assess the natural aerosol particle deposition process in a canister in the event that a release occurs from one or more fuel pins. In the previous work it was demonstrated that CFD can model the flow through a PWR spacer grid with mixing vanes, including particle deposition, in a reasonable amount of time on a personal computer. In this work, the analysis is extended to include the bypass region between an assembly and the canister basket walls. It is shown that the flow velocity in the bypass region is about three times that of the interstitial region between the pins. The lengths before and after the spacer grid are also extended to determine when the flow becomes fully developed. In addition, the approach of computationally “stitching together” segments of an assembly is demonstrated with the plan to ultimately model a full assembly. The fraction of particles that are deposited in a segment with a spacer grid is determined as a function of particle size and flow velocity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Specifications of Cladding Diameter Measurements Conducted at AGHCF

Fuel element diameter data was collected on-site at the Alpha-Gamma Hot Cell Facility (AGHCF) before and after out-of-pile furnace transient tests of fuel elements. Available data records have been collected and preserved in the Out-of-Pile Transient Database (OPTD). This data is used to determine the transient-induced changes in fuel element diameter, or cladding strain, for the Whole Pin Furnace (WPF) tests. Fuel element diameter was measured by contact profilometry along the length of the fuel pin at specified rotational orientations and/or by a manually operated micrometer at several discrete axial locations along the pin. The Alpha-Gamma Hot Cell Facility Operations Manual details the way the facility operated, organizational and oversight responsibilities, and procedures for examination of samples. The working version of the manual at the time of the whole pin furnace tests and the test pin examinations is Doc. No. IPS-2-00-00, dated June 1989. This specification was developed using the operations manual and recovered measurement records in consultation with subject matter experts (SMEs). The measurement methods, format of the available post-test examination (PTE) data, and recommended methods for use and interpretation of that data are summarized. Section 2 describes the available diameter measurement data for the WPF tests with guidance for interpretation and usage, and Section 3 describes the instrument measurement procedures and calibration methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

DASSH-F: Subchannel Based Thermal Analysis

The DASSH thermal analysis code is designed to rapidly allow a reactor design engineer to obtain flow rates requirements that satisfy peak temperature constraints in the domain. The advantage of using DASSH over a hand calculation is that it has a more rigorous treatment of the pin power distribution and coolant heat transfer within an assembly and between assemblies. The advantage of using DASSH over a conventional 3D subchannel code or a computational fluid dynamics code (CFD) is that it can obtain the desired solution in a matter of minutes in serial with minor computer memory needs. The DASSH methodology for pin lattice models is virtually identical to SUPERENERGY-2 with additional functionalities taken from follow on work to SUPERENERGY-2 done at ANL in the 1980s. DASSH today is an integral component of the Argonne Fast Reactor analysis suite for reactor design work. DASSH obtains the power distribution from a coupled neutron-gamma heating calculation in GAMSOR (including DIF3D) at each time point of a companion fuel cycle analysis calculation with REBUS. The domain in DASSH assumes a hexagonal grid typical for fast reactors with much of the geometry information taken from the DIF3D model. DASSH assumes the assemblies that are loaded into each grid position are ducted to control the coolant flow. The user can alternatively provide their own geometry and power profile instead of inheriting it from DIF3D. Considerable detail is given on the subchannel formulation of DASSH in this document. Much of the formulation and design of the code builds upon research done by previous authors with little new investigation. Thus the decisions made in developing the subchannel model used in DASSH have their origins over 50 years ago. Much of the heat transfer methodology in DASSH is built upon correlations for both the coolant mixing and heat transfer coefficients for pins and ducts. DASSH is thus not a rigorous treatment of a given problem, but a rapid assessment of the temperature field that has known limitations with respect to an experimental measurement or CFD calculation. The DASSH input and output are detailed along with usage of the software. The DASSH output provides tables of evaluated material properties and key coolant and pin temperature results. DASSH can create Python scripts that generate domain summary pictures. DASSH can also generate assembly temperature maps and VTK output files which allow the DASSH solution to be visualized. As the primary purpose of the DASSH software is to compute the coolant and fuel pin temperature distribution for a given model of a reactor, much of the output focus is giving the user quick summary tables needed to assess the performance of a given orifice flow specification. The present version of DASSH has a crude orifice search capability and a sufficient orifice flow search capability. The flow search tries to meet user specified constraints for 1) peak 2-sigma clad temperature, 2) peak coolant temperature, and 3) desired bulk outlet temperature. This document serves as the manual for the Fortran based DASSH software that was developed to replace the Python version of DASSH developed as part of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1971

Molybdenum-UO2 cerment irradiation at 1145 K

Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

Mcdonald, G.↗
NASA NTRS1982

Thermally Insulated Glove With Good Tactility

Thermally insulated glove contains short, closely-spaced elastomeric pins that insulate without impairing flexibility. By confining pins to the inter-joint areas of palm, fingers and back of the hand, joint mobility is retained. Glove thermal-insulation requirements dictate the relationships among pin length, pin diameter, and number of pins per unit surface length.

Balinskas, R.↗
NASA NTRS1984

Vortex creep and the internal temperature of neutron stars. II - Vela pulsar

The observed complex postglitch behavior of the Vela pulsar is explained as resulting from coupling of the crust to crustal neutron superfluid, specifically that part of the superfluid in which vortex lines are pinned to crustal nuclei. It is shown how the general theory of vortex creep provides an excellent fit to the timing observations of Downs which span the decade 1969-1979 and include four giant glitches. Relaxation times, inertial moments, and limits on superfluid pinning parameters are extracted for three distinct regions of vortex pinning in the star, with results which are consistent with microscopic theories of its internal structure. Relaxation times due to vortex creep are directly proportional to the internal temperature of the star, so that the limits obtained for pinning parameters translate to bounds on this temperature. It is concluded that the internal temperature of the Vela pulsar is about 10-million K and discuss the extent to which improved calculations of vortex pinning as well as soft X-ray observations of other stars will make possible an improved determination of the pulsar temperature.

Alpar, M. A.↗
NASA NTRS1984

T/R switch design for short-range measurements, part 6.1A

The positive intrinsic negative (PIN) diode switch which is designed to protect the receiver from burnout or damage on transmission and channel the echo signal to the receiver on reception is outlined. The receiver must be protected firmly. A schematic diagram of a transformer rectifier (TR-ATR) switch for the Urbana Radar is shown. The T/R switch consists of a half wavelength coaxial cavity with tuning condenser and PIN diodes. Two UM4300 PIN diodes were mounted between the inner and outer conductor. The dc biasing voltage required for the PIN diodes is supplied by a control circuit. On transmission, the PIN diodes are forward biased to about 0.5 amperes. On reception, about 10 volts reverse voltage is applied to the diodes, which produces an initial reverse current to speed the recovery time. The T/R switch characteristics are estimated and the result of testing at different peak transmitter powers from 410 kW to 1500 kW is shown.

Yu, B.↗
NASA NTRS1990

Double-Swivel Mechanism For Reliable Release

Double-swivel toggle mechanism releases large, heavy objects reliably. Double-swiveling action of mechanism ensures it clears restraining pins upon release. Pins retain toggle and its load. If pin fails to withdraw at designated time for releasing payload, toggle swivels about its upper ball, and ring swivels about lower ball so ring flange clears failed pin. Double-swivel action ensures disengagement even it two pins fail to withdraw.

King, Guy L.↗
NASA NTRS1989

Double swivel toggle release

A pyrotechnic actuated structural release device is disclosed which is mechanically two fault tolerant for release. The device comprises a fastener plate and fastener body each attachable to one of a pair of structures to be joined. The fastener plate and the fastener body are fastened by a dual swivel toggle member. The toggle member is supported at one end on the fastener plate and mounted for universal pivotal movement thereon. Its other end is received in a central opening in the fastener body, and has a universally mounted retainer ring member. The toggle member is restrained by three retractable latching pins symmetrically disposed in equiangular spacing about the axis of the toggle member and positionable in latching engagement with the retainer ring member on the toggle member. Each pin is retractable by a pyrotechnic charge, the expanding gases of which are applied to a pressure receiving face on the latch pins to effect retraction from the ring member. While retraction of all three pins releases the ring member, the fastener is mechanically two fault tolerant since the failure of any single one or pair of the latch pins to retract results in an asymmetrical loading on the ring member and its dual pivotal movement ensures a release.

King, Guy L.↗
NASA NTRS1991

Thermal shock testing for assuring reliability of glass-sealed microelectronic packages

Tests were performed to determine if thermal shocking is destructive to glass-to-metal seal microelectronic packages and if thermal shock step stressing can compare package reliabilities. Thermal shocking was shown to be not destructive to highly reliable glass seals. Pin-pull tests used to compare the interfacial pin glass strengths showed no differences between thermal shocked and not-thermal shocked headers. A 'critical stress resistance temperature' was not exhibited by the 14 pin Dual In-line Package (DIP) headers evaluated. Headers manufactured in cryogenic nitrogen based and exothermically generated atmospheres showed differences in as-received leak rates, residual oxide depths and pin glass interfacial strengths; these were caused by the different manufacturing methods, in particular, by the chemically etched pins used by one manufacturer. Both header types passed thermal shock tests to temperature differentials of 646 C. The sensitivity of helium leak rate measurements was improved up to 70 percent by baking headers for two hours at 200 C after thermal shocking.

Thomas, Walter B., III↗
NASA NTRS1992

Simple, Lightweight, Thermomechanical Coupling

Self-actuating thermal coupling device is light, and transfers heat efficiently. Coupling is thermostatic, making contact as necessary to maintain nearly steady temperature at hot side of interface. Hot pins on hot plate poised near cold pins on cold-plate. As temperature of hot plate rises, shape-memory wire contracts, drawing hot pin against cold pin. When temperature falls below critical level, wire expands, allowing hot pin to pull away. Critical temperature determined by composition of alloy.

Milam, M. Bruce↗
NASA NTRS1992

Environmental effects on friction and wear of diamond and diamondlike carbon coatings

Reciprocating sliding friction experiments were conducted with a natural diamond flat, diamond film, and low and high density diamondlike carbon (DLC) films in contact with pin specimens of natural diamond and silicon nitride (Si3N4) both in humid air and dry air nitrogen. The results indicated that for natural diamond pin contacts the diamond films and the natural diamond flat were not susceptible to moisture but that moisture could increase both the coefficient of friction and the wear factors of the DLC films. The coefficients of friction and wear factors of the diamond films were generally similar to those of the natural diamond flat both in humid air and dry air nitrogen. In dry nitrogen the coefficients of friction of the high density DLC films in contact with pin specimens of both diamond and Si3N4 were generally low (about 0.02) and similar to those of the natural diamond flat and the diamond films. The wear factors of the materials in contact with both natural diamond and Si3N4 were generally in the ascending order of natural diamond flat, diamond film, high density DLC film, and low density DLC film. The moisture in the environment increased the coefficients of friction for Si3N4 pins in contact with all the materials. This increase in friction is due to the silicon oxide film produced on the surface of Si3N4 pins in humid air.

Miyoshi, Kazuhisa↗
NASA NTRS1999

Fundamental Study of Material Flow in Friction Stir Welds

The presented research project consists of two major parts. First, the material flow in solid-state, friction stir, butt-welds as been investigated using a marker insert technique. Changes in material flow due to welding parameter as well as tool geometry variations have been examined for different materials. The method provides a semi-quantitative, three-dimensional view of the material transport in the welded zone. Second, a FSW process model has been developed. The fully coupled model is based on fluid mechanics; the solid-state material transport during welding is treated as a laminar, viscous flow of a non-Newtonian fluid past a rotating circular cylinder. The heat necessary for the material softening is generated by deformation of the material. As a first step, a two-dimensional model, which contains only the pin of the FSW tool, has been created to test the suitability of the modeling approach and to perform parametric studies of the boundary conditions. The material flow visualization experiments agree very well with the predicted flow field. Accordingly, material within the pin diameter is transported only in the rotation direction around the pin. Due to the simplifying assumptions inherent in the 2-D model, other experimental data such as forces on the pin, torque, and weld energy cannot be directly used for validation. However, the 2-D model predicts the same trends as shown in the experiments. The model also predicts a deviation from the "normal" material flow at certain combinations of welding parameters, suggesting a possible mechanism for the occurrence of some typical FSW defects. The next step has been the development of a three-dimensional process model. The simplified FSW tool has been designed as a flat shoulder rotating on the top of the workpiece and a rotating, cylindrical pin, which extends throughout the total height of the flow domain. The thermal boundary conditions at the tool and at the contact area to the backing plate have been varied to fit experimental data such as temperature profiles, torque and tool forces. General aspects of the experimentally visualized material flow pattern are confirmed by the 3-D model.

Reynolds, Anthony P.↗
NASA NTRS2003

Self-Reacting Friction Stir Welding for Aluminum Alloy Circumferential Weld Applications

Friction stir welding is an innovative weld process that continues to grow in use, in the commercial, defense, and space sectors. It produces high quality and high strength welds in aluminum alloys. The process consists of a rotating weld pin tool that plasticizes material through friction. The plasticized material is welded by applying a high weld forge force through the weld pin tool against the material during pin tool rotation. The high weld forge force is reacted against an anvil and a stout tool structure. A variation of friction stir welding currently being evaluated is self-reacting friction stir welding. Self-reacting friction stir welding incorporates two opposing shoulders on the crown and root sides of the weld joint. In self-reacting friction stir welding, the weld forge force is reacted against the crown shoulder portion of the weld pin tool by the root shoulder. This eliminates the need for a stout tooling structure to react the high weld forge force required in the typical friction stir weld process. Therefore, the self-reacting feature reduces tooling requirements and, therefore, process implementation costs. This makes the process attractive for aluminum alloy circumferential weld applications. To evaluate the application of self-reacting friction stir welding for aluminum alloy circumferential welding, a feasibility study was performed. The study consisted of performing a fourteen-foot diameter aluminum alloy circumferential demonstration weld using typical fusion weld tooling. To accomplish the demonstration weld, weld and tack weld development were performed and fourteen-foot diameter rings were fabricated. Weld development consisted of weld pin tool selection and the generation of a process map and envelope. Tack weld development evaluated gas tungsten arc welding and friction stir welding for tack welding rings together for circumferential welding. As a result of the study, a successful circumferential demonstration weld was produced leading the way for future circumferential weld implementation.

Bjorkman, Gerry↗
NASA NTRS2003

Investigation of Friction Stir Welding of Al Metal Matrix Composite Materials

The innovative process of Friction Stir Welding (FSW) has generated tremendous interest since its inception about a decade or so ago since the first patent in 1991 by TWI of Cambridge, England. This interest has been seen in many recent international conferences and publications on the subject and relevant published literature. Still the process needs both intensive basic study of deformation mechanisms during this FSW process and analysis and feasibility study to evaluate production methods that will yield high quality strong welds from the stirring action of the appropriate pin tool into the weld plate materials. Development of production processes is a complex task that involves effects of material thickness, materials weldability, pin tool design, pin height, and pin shoulder diameter and related control conditions. The frictional heating with rotational speeds of the pin tool as it plunges into the material and the ensuing plastic flow arising during the traverse of the welding faying surfaces provide the known special advantages of the FSW process in the area of this new advanced joining technology.

Diwan, Ravinder M.↗