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At least 1,513 records · Page 84
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

10-160B↗
DOE OSTI2016

Velocimetry Overview for visitors from the DOD

We are in the midst of a transformative period in which technological advances are making fundamental changes in the measurement techniques that form the backbone of nuclear weapon certification. Optical velocimetry has replaced electrical shorting pins in “Hydrotests,” which measure the dynamic implosion process. This advance has revolutionized nuclear weapons certification during the last 5 years. We can now measure the implosion process that drives a nuclear detonation with many orders of magnitude more resolution in both space and time than was possible just 10 years ago. It has been compared to going from Morse Code to HDTV, resulting in a dozen or more improvements in models of these weapons. These Hydrotests are carried out at LANL, LLNL and the NNSS, with the later holding the important role of allowing us to test with nuclear materials, in sub-critical configurations (i.e., no yield.) Each of these institutions has largely replaced pins with hundreds of channels of optical velocimetry. Velocimetry is non-contact and is used simultaneously with the X-ray capability of these facilities. The U1-a facility at NNSS pioneered this approach in the Gemini series in 2012, and continues to lead, both in channel count and technological advances. Close cooperation among LANL, LLNL and NSTec in these advances serves the complex by leveraging capabilities across sites and accelerating the pace of technical improvements.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Complete Status Report Documenting Weld Development for Thin Wall Tubing of ODS Ferritic Alloys

Beginning in 2015, research in the FCRD program began the development of FSW for joining thin sections of 14YWT in the form of thin (0.5 mm) plate and ultimately thin wall tubing. In the previous fiscal year, a ~1 mm thick plate, or sheet, of 14YWT was produced by hot rolling with no edge cracking. The initial FSW experiment was performed on the 1 mm thick plate and involved a bead-on-plate weld in which the spinning pin tool is plunged into the plate surface, but does not penetrate the thickness of the plate, and then travels the length of the plate. The FSW run successfully produced a bead-on-plate stir zone on the 1 mm thick plate of 14YWT, but no characterization studies of the stir zone were performed by the end of FY15. Therefore, the results presented in this report cover the microstructural analysis of the bead-on-plate stir zone and the initial research task on obtaining tensile properties of the stir zone using the digital image correlation (DIC) approach during testing of miniature tensile specimens to assess the quality of the FSW parameters used in the initial experiment. The results of the microstructural characterization study using optical, scanning electron and scanning transmission electron microscopies showed the grain structure in the SZ to have isotropic and irregular shape but very similar size compared to the highly elongated grains oriented horizontally with the plane of the plate that were observed in the unaffected zone of 14YWT. Several cracks oriented horizontally were observed mostly on the retreating side of the SZ in both the SZ and TMAZ. These cracks may have formed due to insufficient pressure being exerted on the top surface of the plate by the shoulder and pin tool during the FSW run. High resolution STEM-EDS analysis showed the presence of the Y-Ti-O particles in the SZ, but that some particles exhibited coarsening. Overall, the FSW parameters used to produce the bead-on-plate SZ in the 0.1 cm thick plate of 14YWT were nearly optimized. The results of the digital image correlation (DIC) analysis of the two SS-Mini-2 tensile specimens fabricated from the 0.1 cm thick plate of 14YWT showed that the specimens exhibited high strength and good ductility. However, strain localization occurred in one of the specimens during the tensile test that was too close to the grips, which invalidated the data from the DIC analysis. This was surprising since the abrupt crack pop-in that occurred in the 0.1 cm thick plate of 14YWT during fabrication by wire EDM suggested that residual stresses were high. Residual stress measurements and the effects of post weld heat treatment on the FSW quality of joined 14YWT plates will be investigated in the next FW work package.

36 MATERIALS SCIENCE↗
DOE OSTI2016

MPACT Theory Manual, Version 2.2.0

This theory manual describes the three-dimensional (3-D) whole-core, pin-resolved transport calculation methodology employed in the MPACT code. To provide sub-pin level power distributions with sufficient accuracy, MPACT employs the method of characteristics (MOC) solutions in the framework of a 3-D coarse mesh finite difference (CMFD) formulation. MPACT provides a 3D MOC solution, but also a 2D/1D solution in which the 2D planar solution is provided by MOC and the axial coupling is resolved by one-dimensional (1-D) lower order (diffusion or P3) solutions. In Chapter 2 of the manual, the MOC methodology is described for calculating the regional angular and scalar fluxes from the Boltzmann transport equation. In Chapter 3, the 2D/1D methodology is described, together with the description of the CMFD iteration process involving dynamic homogenization and solution of the multigroup CMFD linear system. A description of the MPACT depletion algorithm is given in Chapter 4, followed by a discussion of the subgroup and ESSM resonance processing methods in Chapter 5. The final Chapter 6 describes a simplified thermal hydraulics model in MPACT.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of korzincalloy prepared by Hohman Plating

A commercial vendor, Hohman Plating performed contract engineering work to determine the feasibility of producing pin hole free KorZincAlloy bronze material used for zinc gettering. Samples were tested for Sn plating thickness, heat treatability, and chemistry prior to being subjected to a standardized zinc exposure. The samples absorbed zinc and were examined using visual and scanning electron microscopy. Hohman Plating successfully produced KZA that met the target composition, was pin hole free, and was an effective zinc getter.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Warthog: Coupling Status Update

The Warthog code was developed to couple codes that are developed in both the Multi-Physics Object-Oriented Simulation Environment (MOOSE) from Idaho National Laboratory (INL) and SHARP from Argonne National Laboratory (ANL). The initial phase of this work, focused on coupling the neutronics code PROTEUS with the fuel performance code BISON. The main technical challenge involves mapping the power density solution determined by PROTEUS to the fuel in BISON. This presents a challenge since PROTEUS uses the MOAB mesh format, but BISON, like all other MOOSE codes, uses the libMesh format. When coupling the different codes, one must consider that Warthog is a light-weight MOOSE-based program that uses the Data Transfer Kit (DTK) to transfer data between the various mesh types. Users set up inputs for the codes they want to run, and then Warthog transfers the data between them. Currently Warthog supports XSProc from SCALE or the Sub-Group Application Programming Interface (SGAPI) in PROTEUS for generating cross sections. It supports arbitrary geometries using PROTEUS and BISON. DTK will transfer power densities and temperatures between the codes where the domains overlap. In the past fiscal year (FY), much work has gone into demonstrating two-way coupling for simple pin cells of various materials. XSProc was used to calculate the cross sections, which were then passed to PROTEUS in an external file. PROTEUS calculates the fission/power density, and Warthog uses DTK to pass this information to BISON, where it is used as the heat source. BISON then calculates the temperature profile of the pin cell and sends it back to XSProc to obtain the temperature corrected cross sections. This process is repeated until the convergence criteria (tolerance on BISON solve, or number of time steps) is reached. Models have been constructed and run for both uranium oxide and uranium silicide fuels. These models demonstrate a clear difference in power shape that is not accounted for in a stand-alone BISON run. Future work involves improving the user interface (UI), likely through integration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Workbench. Furthermore, automating the input creation would ease the user experience. The next priority is to continue coupling the work with other codes in the SHARP package. Efforts on other projects include work to couple the Nek5000 thermo-hydraulics code to MOOSE, but this is in the preliminary stages.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

In-situ X-Ray Analysis of Rapid Thermal Processing for Thin-Film Solar Cells: Closing the Gap between Production and Laboratory Efficiency

For materials synthesis, it is well known that the material final state may not reach equilibrium and depends on the synthetic process. In particular, processes that quickly remove the available energy from the material may leave it in a metastable state and the metastability may actually impart desirable functional properties. By its very nature, Rapid thermal processing (RTP) is ideally suited to produce such metastable materials. However, metastability and the dynamics of reaching a metastable state are poorly understood, since this is best accomplished through in situ monitoring. In this regard, RTP is particularly challenging as the processing time are very short (seconds to minutes). As a result, there is only poor understanding, and hence use, of RTP in industry. This is potentially a cost-increasing limitation, because RTP can decrease cost by decreasing processing time, and as such, increase throughput and decrease the total thermal budget of processing - a significant cost. RTP is already being used for key processing steps in PV technologies. With silicon wafer PV, it is used for establishing electrical contact between the Ag metal grid and the silicon (known as firing). In this process, a silicon wafer with deposited metal/frit in a grid pattern is heated rapidly to temperatures between 750 and 800 ºC. The processing time when the temperature is held above 600ºC is short (<5 seconds). This process has historically been optimized empirically and it is unclear how the thermal processing affects formation of the final contact between the metal and the silicon. In the case of thin-film PV, RTP has been demonstrated in the process of making absorber layers, i.e. CIGS and CZTS. Use of RTP can reduce the processing time from 10s of minutes to seconds, reducing the thermal budget and increasing the throughput significantly. The conversion from precursor material to final PV material is not well understood, and most of the process optimization is done empirically. Again, the proposed in situ monitoring will provide insight into the conversion process. We have developed a unique instrument for in situ monitoring of the reaction process during RTP. The most fundamental property to monitor is the atomic arrangement (for solids, crystalline structure) and thus we have focused on this property. The chamber was designed to collect X-ray diffraction (XRD) for crystal phase determination, although it has lent insight in the formation of liquids. We designed and constructed both inert (air) and reactive (S/Se) versions of this chamber. We have applied the RTP-XRD methodology to: silicon solar cell electrical contact formation; the reaction sequence for forming CIGS PV thin films from copper selenide and indium gallium selenide precursors; the phase sequence for forming HC(NH2)2PbI3 (FAPbI3) and CH3NH3PbI3 (MAPbI3) hybrid-perovskite PV thin films. The RTP instrument enables real-time collection of X-ray diffraction data with intervals as short as 100 ms, and while heating at ramp rates as high 100 ºCs-1 up to 1200 ºC. The system is portable and can be installed on a synchrotron beamline. The RTP reactor consists of a metallic body (front cap assembly, central block and end cap assembly) that houses a quartz reaction chamber. The sample is secured inside the reaction chamber on a sample holder connected to the front-cap assembly. Modifications of the chamber were made for safely handling sulfur and selenium vapor environments. A Se/S source holder was added to generate a Se/S over pressure in the chamber on heating and the entrance window is made wider in order to accommodate the detachable additional holder. Se/S are toxic in nature and therefore a proper Se/S trapping is used before releasing the gases to exhaust. Screen-printing provides an economically attractive means for making Ag electrical contacts to Si solar cells, but the use of Ag substantiates a significant manufacturing cost, and the glass frit used in the paste to enable contact formation contains Pb. To achieve optimal electrical performance and to develop pastes with alternative, abundant, and non-toxic materials requires understanding the contact formation process during firing. We use the RTP-XRD chamber during firing of Ag-Si contacts to reveal the reaction sequence for contact formation. The findings show that between 500 ºC and 650 ºC PbO in the frit etches the SiNx antireflective-coating on the solar cell, exposing the Si surface. Then, above 650 ºC, Ag+ dissolves into the molten glass frit – key for enabling deposition of metallic Ag on the emitter surface and precipitation of Ag nanocrystals within the glass. This clarifies contact formation mechanisms and can be used to suggest approaches for development of inexpensive, nontoxic solar cell contacting pastes. We used the RTP/XRD chamber to study the formation of CuInGaS/Se (CIGS) PV films by in-situ selenization of pre-deposited metal stacks and annealing of pre-deposited Cu-In-Ga-Se precursors. During the metal stack selenization, we observed that Se first reacts with the metals and forms selenides. We observed formation of intermetallics such as Cu2In and binary selenides during heating. On further annealing in a Se atmosphere, CuInSe2 and CuGaSe2 forms which on heating up to 550 °C reacts further to form the CIGS phase. On prolonged exposure to 550 °C for more than 5 mins or higher temperature, we see separation into Ga-rich and Ga-depleted CIGS phases. On introducing a holding step at 400 °C for 10 min during the annealing ramp, the CIGS phase (forms between 500-550 °C), appears to be less prone to decomposition into high-Ga and low-Ga phases during holding at peak temperature. Lead halide perovskites have emerged as successful optoelectronic materials with high photovoltaic power conversion efficiencies and low material cost. However, substantial challenges remain in the scalability, stability and fundamental understanding of the materials. We demonstrated the application of radiative thermal annealing, an easily scalable processing method, for synthesizing formamidinium lead iodide (FAPbI3) perovskite solar absorbers. Devices fabricated from films formed via radiative thermal annealing have equivalent efficiencies to those annealed using a conventional hotplate. By coupling results from in situ x-ray diffraction using a radiative thermal annealing system with device performances, we mapped the processing phase space of FAPbI3 and corresponding device efficiencies. Our map of processing-structure-performance space suggests the commonly used FAPbI3 annealing time, 10 min at 170 °C, can be significantly reduced to 40 s at 170 °C without affecting the photovoltaic performance. Applying the Johnson-Mehl-Avrami model to the decomposition into PbI2, the activation energy was determined. We also presented a well-controlled, manufacturing relevant annealing method, radiative thermal annealing, for the methylammonium lead triiodide films formed by solvent engineering process, with dimethylformamide (DMF) and dimethyl sulfoxide as solvent and diethyl ether as the antisolvent. Radiative thermal annealing can produce high quality perovskite films, evidenced by high efficiency solar cell devices, in a shorter time than the widely-used hotplate annealing. Using in-situ x-ray diffraction during the radiative annealing, we show that the role of the antisolvent is not to form an intermediate compound (a methylammonium iodide, lead iodide, and dimethyl sulfoxide compound) by washing of the main solvent (DMF) but to achieve a pin-hole free, uniform film of methylammonium lead iodide. We find that having the intermediate compound does not guarantee a high quality (pin-hole free) perovskite film. Our study was extended to reveal the effect of annealing environment, annealing temperature profile, and deposition substrate.

36 MATERIALS SCIENCE↗
DOE OSTI2017

FY 2017 Modeling of the M8 Calibration Series using MAMMOTH

This report provides an overview of the modeling efforts performed in FY17 to validate MAMMOTH using historical TREAT data. The data for comparison comes from the M8 series of calibration measurements, which was the last series performed before TREAT was shut down in 1994. The data set includes: an experiment flux profile; low level steady state power coupling factors for flux wires and fuel pins; and a temperature limited transient. MAMMOTH was able to obtain reasonable results for all cases with the exception of the H91-8-1 flux wire. Because of issues encountered in determination of biases introduced in measurement of power coupling factors, there was insufficient time to model the fuel pins in MAMMOTH. However, the feasibility of doing so was demonstrated with the flux wire calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

MOOSE Implementation of MAMBA

The development of MAMBA is targeted at capturing both core wide CRUD induced power shifts (CIPS) as well as pin-­level CRUD induced localized corrosion (CILC). Both CIPS and CILC require some sort of information from thermal-­hydraulic, neutronics, and fuel performance codes, although the degree of coupling is different for the two effects. Since CIPS necessarily requires a core-­wide power distribution solve, it requires tight coupling with a neutronics code. Conversely, CIPS tends to be an individual pin phenomenon, requiring tight coupling a fuel performance code. As efforts are now focused on coupling MAMBA within the VERA suite, a natural separation has surfaced in which a FORTRAN rewrite of MAMBA is optimal for VERA integration to capture CIPS behavior, while a CILC focused calculation would benefit from a tight coupling with BISON, motivating a MOOSE version of MAMBA.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Industrializing Offshore Wind Power with Serial Assembly and Lower-cost Deployment - Final Report

A team of engineers and contractors has developed a method to move offshore wind installation toward lower cost, faster deployment, and lower environmental impact. A combination of methods, some incremental and some breaks from past practice, interact to yield multiple improvements. Three designs were evaluated based on detailed engineering: 1) a 5 MW turbine on a jacket with pin piles (base case), 2) a 10 MW turbine on a conventional jacket with pin piles, assembled at sea, and 3) a 10 MW turbine on tripod jacket with suction buckets (caissons) and with complete turbine assembly on-shore. The larger turbine, assembly ashore, and the use of suction buckets together substantially reduce capital cost of offshore wind projects. Notable capital cost reductions are: changing from 5 MW to 10 MW turbine, a 31% capital cost reduction, and assembly on land then single-piece install at sea an additional 9% capital cost reduction. An estimated Design 4) estimates further cost reduction when equipment and processes of Design 3) are optimized, rather than adapted to existing equipment and process. Cost of energy for each of the four Designs are also calculated, yielding approximately the same percentage reductions. The methods of Design 3) analyzed here include accepted structures such as suction buckets used in new ways, innovations conceived but previously without engineering and economic validation, combined with new methods not previously proposed. Analysis of Designs 2) and 3) are based on extensive engineering calculations and detailed cost estimates. All design methods can be done with existing equipment, including lift equipment, ports and ships (except that design 4 assumes a more optimized ship). The design team consists of experienced offshore structure designers, heavy lift engineers, wind turbine designers, vessel operators, and marine construction contractors. Comparing the methods based on criteria of cost and deployment speed, the study selected the third design. That design is, in brief: a conventional turbine and tubular tower is mounted on a tripod jacket, in turn atop three suction buckets. Blades are mounted on the tower, not on the hub. The entire structure is built in port, from the bottom up, then assembled structures are queued in the port for deployment. During weather windows, the fully-assembled structures are lifted off the quay, lashed to the vessel, and transported to the deployment site. The vessel analyzed is a shear leg crane vessel with dynamic positioning like the existing Gulliver, or it could be a US-built crane barge. On site, the entire structure is lowered to the bottom by the crane vessel, then pumping of the suction buckets is managed by smaller service vessels. Blades are lifted into place by small winches operated by workers in the nacelle without lift vessel support. Advantages of the selected design include: cost and time at sea of the expensive lift vessel are significantly reduced; no jack up vessel is required; the weather window required for each installation is shorter; turbine structure construction is continuous with a queue feeding the weather-dependent installation process; pre-installation geotechnical work is faster and less expensive; there are no sound impacts on marine mammals, thus minimal spotting and no work stoppage Industrializing Offshore Wind Power 6 of 96 9 for mammal passage; the entire structure can be removed for decommissioning or major repairs; the method has been validated for current turbines up to 10 MW, and a calculation using simple scaling shows it usable up to 20 MW turbines.

17 WIND ENERGY↗
DOE OSTI2017

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO 2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO 2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave similarly and retain the same ex-core features and characteristics. While INL has developed Designs A and B, LANL is evaluating a Hot Isostatic Pressed reactor configuration to overcome the drawbacks of the monolithic core design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

20 Channel Quench Detector

A multi-channel quench detector is useful for troubleshooting of prototype HTC superconducting magnets with current lead connections and imbedded splices. There are commercially available quench detectors, but not with as many channels, or with the particular functionality of this unit. This unit was designed specifically for cold testing of a superconducting sextupole prototype, which is wound with many turns of HTC ribbon. Associated with the current lead connections and the imbedded splices are 18 voltage taps, thus 17 quench detection channels are required to monitor the condition of the magnet. Of the 20 differential input channels in this quench detection unit, 19 are configured in a ladder network. The 20th channel is isolated and can be connected across any pair of voltage taps. This quench detector is designed to be used to monitor the condition of a series-connected superconducting magnet powered by a unipolar supply, and therefore the polarity and sequence of the voltage taps must be observed. There are 20 toggle switches and 20 red LEDs on the front panel of the unit which are associated with the 20 quench detection channels. There are also 2 green LEDs to indicate the state of the +5 V and +25 V internal supplies. Channel input impedance is ~10 kΩ. The input lines are diode protected against negative voltage transients and can tolerate up to +25 V, which is greater than the compliance of supplies typically used to power superconducting magnets. The front panel switches are used to enable or disable individual channels. If a channel is not used it is disabled. There are 20 pots on the quench detector PCB. They provide adjustment for the sensitivity of each of the 20 channels. There is an onboard reference to facilitate channel sensitivity calibration. The detection sensitivity range is from ~27 mV to ~700 mV. Typically all channels are set for the same sensitivity. The response time is ~20 μs, but the time constant can be increased if desired by adding a single capacitor to each channel. When the magnet under test is cooled to the superconducting state and the supply current to the windings is slowly ramped up to the set point, if there is a voltage rise across a pair of voltage taps which exceeds the quench detection limit of a given channel, then the front panel LED for that channel illuminates and all other channels are disabled within ~25 ns. Thus only the first quench is detected for the channels which are enabled. A pushbutton switch on the front panel is used to reset a quench detection. A 25 pin D-subminiature connector on the back of the unit is used for the channel inputs. TTL level ‘quench’ and ‘quench-not’ signals, and DPDT relay contacts, are provided through a 9 pin D-subminiature connector on the back to indicate a detected quench condition, which could be used to disable the power supply and engage an energy absorber.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Complete Status Report Documenting the Development of Friction Stir Welding for Producing a Butt Joint in Thin Wall Tubing of ODS Alloys

A challenging attempt using friction stir welding to produce a butt joint between two 1.0 mm thick plates of 14YWT was nearly successful. The modified tool pin mostly penetrated the thickness of the adjoined thin plate samples resulting in a small blunt notch on the bottom surface of the butt joint. A bend occurred in the thin plate sample due to formation of stresses from clamping the thin plate sample during FSW. The peak displacement of the bend occurred along the center line of the butt joint. The weld zone showed good microstructure mixing with some flashing mostly on the retreating side that suggested excess heat input by friction between the traveling and spinning tool pin and thin plates of 14YWT. Neutron diffraction experiments were performed on the bent sample of 14YWT to measure the residual stresses in the butt joint on the HB-2B beam line at the Neutron Residual Stress Facility at HFIR. However, the neutron scattering to background noise ratio was very low during data acquisition. Measurements of lattice strains were obtained in the rolling (RD) and normal (ND) directions but not in the extrusion direction (ED) of the thin plate sample. Unfortunately, the quality of the data was not sufficient for determining the residual stresses in the butt joint. From the previous neutron scattering experiments conducted on four hot rolled plates (R1, R2, R3 and R4) of 14YWT, the residual stresses were calculated from lattice strain data and plotted in 2-D color contour maps for ED, RD and ND. The results showed variations in the magnitude and spatial position of residual stresses in the microstructure of the four plates. The magnitude in residual stresses in ED and RD increased gradually with increasing rolling deformation for R1 (50%) to R2 (76%), followed by larger increases in R3 (89%) and R4 (95%). The maximum peaks in residual stresses for R4 was ~710-730 MPa in ED and RD. The residual stresses in ND were below ~300 MPa for each of the 4 rolled plates, with a maximum residual stress in R4 (95%) of ~400 MPa. These results indicated that significant residual stresses representing ~57% of the yield stress of 14YWT (SM13) had accumulated in the 14YWT plate after hot rolling to the 1 mm thickness, which was used in the FSW experiments. The microstructure analysis of the 1.0 mm thick plate of 14YWT containing the butt joint produced by FSW revealed significant changes in the grain structure and texture between the unaffected zone (UZ) of the rolled plate and the butt joint weld. In the UZ, the grains were severely elongated after 95% deformation and mainly possessed the {101} texture component in the ED. Two predominant grain misorientation angles were observed: one consisting of small misorientations of <5º between highly elongated grains and the other consisting of large misorientations of >15º in pockets of small grains that might have formed by dynamic recrystallization processes during rolling of the plate at 1000ºC. The morphology and texture of grains in the heat affected zone (HAZ) were similar to those in UZ, except the grains had coarsened due to excess heat input during FSZ. A population of pores formed and the grains both coarsened and changed to isotropic morphology in the microstructure of the butt joint on both the advancing and retreating sides. The texture of grains in both regions were randomly orientated between the 3 poles {001}, {101} and {111}. The main difference in texture of grains between the advancing side and retreating side was with the misorientation angles of neighboring grains. Most of the grains in the advancing side had misorientation angles of >5º indicating that these grains experienced recrystallization. On the retreating side, areas containing larger grains showed high misorientation angles of >15º compared to areas appearing as bands of grains with misorientation angles of <5º and between 5º-15º. Numerous small pockets of grains with low misorientation angles of <5º were distributed in the microstructure among the grains showing high misorientation angles of >15º. It was determined that the grains associated with low misorientation angles contained sub-grain structures, which was consistent with both recrystallization and recovery mechanisms for grains located on the retreating side of the butt joint. These results demonstrate the complex microstructure that evolves in the microstructure of 14YWT resulting from the input of high temperatures and severe deformations during FSW.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Steady-State Thermal-Hydraulic Analysis and Bowing Reactivity Evaluation Methods Based on Neutron and Gamma Transport Calculations

With the advancement of computer technology, the variational nodal transport code VARIANT is now routinely used for sodium-cooled fast reactor (SFR) design analysis. In particular, VARIANT transport calculations are usually performed for whole core neutronics analyses and for fuel cycle analyses with REBUS-3. However, the steady-state thermal hydraulics analysis with the SE2-ANL coed and the assembly bowing reactivity calculation with the structural analysis with the NUBOW-3D code are still performed with the DIF3D finite difference diffusion theory code. Furthermore, the DIF3D diffusion calculation for these applications is performed with only six triangular meshes per hexagonal assembly, and the SE2-ANL code determines the pin power distributions approximately by assuming that the power distribution within an assembly is separable in the radial and axial directions. In order to overcome the limitations and to improve the accuracy of the existing methods, it was proposed to develop the relevant computational methods based on the transport calculations with the VARIANT and PROTEUS-SN codes. The objectives of this study are to develop a new heating calculation method based on the coupled neutron and gamma transport calculations with VARIANT, to implement the new heating calculation method into the SE2-ANL steady-state thermal-hydraulics analysis code with improved numerical algorithms and additional capabilities for automatic flow allocation calculations, and to develop a method to calculate bowing reactivity coefficients based on the VARIANT core solutions and the PROTEUS-SN assembly solutions, which can be used in NUBOW-3D calculations. A new, coupled neutron and gamma heating calculation procedure has been developed based on VARIANT transport calculations. In the new heating calculation procedure, the neutron flux distribution is first determined by solving an eigenvalue transport problem using the VARIANT transport code. Using the calculated neutron flux, the intra-nodal gamma source distributions are calculated to be consistent with the neutron flux distribution with the GAMSOR code. For this, the GAMSOR code has been updated to generate the intra-nodal gamma source distributions in the form of the VARSRC dataset of VARIANT. With this gamma source distribution, the gamma flux distribution is determined by solving a fixed source gamma transport problem using VARIANT again. With the neutron and gamma flux distributions, the heat generation rates are evaluated at individual fuel pins and assembly duct walls, without assuming separable axial and radial profiles for the intra-assembly power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 version of the Oak Ridge National Laboratory fast neutron tomographic imaging system. 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 polyethylene cylinder. Measurements demonstrated detectability of about 3 mm. Further data analysis may show improved detection performance and allow a more precise determination of the spatial resolution. Our current simulation code takes only partial consideration of the DT generator neutron production spot size, with neutron production effectively originating at a single point. Simulations showed that features less than 2 mm should be detectable with an improved generator. Our current ING-27 DT generator has a neutron production spot size of about 3.5 mm, which appears to be the limiting factor at present. Generators with spot sizes as small as 1 mm are commercially available. Higher-power, small–spot size associated particle generators are at the prototype stage and are expected to be commercially available soon.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Nuclear Data and Benchmarking Program: Nuclear Data Performance Assessment for Advanced Reactors

Over the last several Evaluated Nuclear Data File (ENDF)/B releases, many updates have been introduced to the nuclear data libraries that can significantly change computational results. These changes can be particularly important for advanced reactor concepts, which are not as thoroughly investigated as those used in light water reactor (LWR) systems. Therefore, a performance assessment of the ENDF/B libraries using the SCALE 6.2.3 code package was conducted under the auspices of the U.S., Department of Energy Office of Nuclear Energy (DOE-NE). The following systems relevant to the advanced reactor community were chosen as subject for investigation in this report: sodium cooled fast reactor systems, graphite moderated high temperature gas-cooled reactors, and several molten salt reactor models. At first, the similarity of the integral performance of the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries caused by compensating errors in important differential data was demonstrated. In calculations for a shipping container of high-assay low enriched uranium, the data of individual isotopes were systematically swapped between the two ENDF/B libraries. An eigenvalue difference of up to 450 pcm was found due to the use of 235 U and 238 U from one library and 1 H and 16 O from the other library. This clearly demonstrates a cross-correlation between reaction data sets of different isotopes within a library that should be reported in the evaluations. For sodium-cooled fast reactor (SFR) fuel assemblies, significant eigenvalue (k inf ) differences (200– 450 pcm) were found between calculations using the 2011 ENDF/B-VII.1 data and 2018 ENDF/B-VIII.0 data. These differences were mainly caused by updates of 238 U and 239 Pu neutron cross sections. Results from the multigroup (MG) calculations further revealed that the group structure of the applied MG library strongly influences the MG bias due to the importance of an appropriate energy resolution of the resonances in higher energy ranges. The application of previously used MG libraries requires new verification for advanced reactor simulation in comparisons with reference continuous-energy calculations. Nuclear data uncertainty analyses of these SFR systems resulted in eigenvalue uncertainties between 1,400 and 1,800 pcm, which are three to four times higher than corresponding uncertainties in light water reactor (LWR) systems. The main contributor to this uncertainty was found to be inelastic scattering on 238 U, which shows an uncertainty of up to 50% in the fast energy range. Other important contributors are the scattering reactions of 56 Fe and 23 Na which have so far not appeared in LWR analysis. Since the largest contribution to the eigenvalue uncertainty of SFR systems is coming from scattering reactions, it is expected that uncertainties in the angular scattering distributions also have a significant impact. Those uncertainties are, however, not available for the majority of nuclides, and the capabilities to determine sensitivities to this data are currently underdeveloped. For two graphite moderated high temperature gas-cooled reactor (HTGR) benchmarks, only the ENDF/BVII. 1 calculations resulted in consistent eigenvalues (k eff ) with the corresponding experimental measurements. Eigenvalue differences of about 1,000 pcm were found between the 2006 ENDF/B-VII.0 data and the 2011 ENDF/B-VII.1 data, primarily due to an updated carbon capture cross section in the thermal energy range in the later ENDF/B release. The ENDF/B-VIII.0 eigenvalue was larger than the ENDF/B-VII.1 eigenvalue by about 300 pcm mainly due to updates in multiple 235 U neutron cross sections, with offsetting updates in the 238 U cross sections. With ENDF/B-VIII.0, graphite can be modeled as perfect crystal or with two different porosities. The choice of the graphite evaluation can have a significant influence of the eigenvalue. For a HTGR benchmark, maximum eigenvalue differences as high as 650 pcm due to different porosities were observed. More detailed studies of the impact of the porosity are necessary while considering that the graphite porosity can vary between the used materials and changes as a function of neutron fluence. xii The uncertainty of the HTGR eigenvalues due to nuclear data uncertainties was found between 500 and 600 pcm, with the top contributor being the neutron multiplicity of 235 U. A gap in the form of missing uncertainties in graphite thermal scattering data, that might have significant impact on HTGR reactors, was identified. This gap is being investigated by the DOE-NE Nuclear Data and Benchmarking Program, as documented in a separate report The fast spectrum molten salt reactor calculations revealed similar comparisons as for the SFR assemblies. The eigenvalue uncertainties were significantly influenced by inelastic scattering on 238 U, while the impact of angular scattering distributions is unknown. For the graphite and zirconium hydride moderated system, eigenvalue uncertainties of up to 700 pcm were observed, requiring in depth analyses of the cross sections and corresponding uncertainties of the included materials. In general, it was observed that the flux spectra of the various molten salt reactor systems show significant differences, even between a fresh state and a depleted state. The choice of the energy group structure of MG calculations is therefore highly relevant; the applicability of previously used MG libraries needs to be verified. Due the unavailability of pin power measurements of advanced reactor systems, pin power calculations with both ENDF/B-VII.1 and VIII.0 data were performed for a light water reactor and compared to corresponding measurements. Both calculations show good agreement with the measurements. A comparison between the ENDF/B-VII.1 and VIII.0 results revealed small differences, mostly in the range of about 0.2%.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗