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At least 1,531 records · Page 85

Machining Graphite Specimens from Sub- Blocks for Baseline Characterization

The Advanced Reactor Technologies (ART) Baseline Graphite Characterization program at the Idaho National Laboratory (INL) is performing a comprehensive evaluation of inter-billet variations in as-produced nuclear-grade graphite. To accomplish this, over 500 mechanical test specimens are machined from a single billet. A typical graphite billet is approximately 6 feet long by 2 feet in diameter or 2 ft square and weighs approximately 1500 pounds. Initially the full-size billets are pre-sectioned into sub-blocks that can be easily handled and shipped. In this SOW, 44 pre-labeled sub-blocks will be machined into mechanical test specimens by the supplier. These sub-blocks are supplied and shipped by the purchaser to the supplier.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

TRISO PARTICLE FUEL POLISHING PROCESS COMPARISON

The investigation of the TRISO coated particles is important in terms of verification the quality of the nuclear fuel samples before placing them in the reactor core. To determine the degree of damage to the coated particles, the samples must be properly polished in order to uncover the inner layers. The poster presents a comparison of two polishing methods. Mechanical polishing method, which is used frequently for sample preparations generates additional mechanical damages in the studied fuel particle, thus directly affecting the experimental results. Due to that, a novel way to polish the samples is presented - an ionic polishing method. Based on the obtained results, it was concluded that the ion polishing method is better because the level of interference with the structures of the individual layers of the tested samples is much lower than with the mechanical method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Insoluble coatings for Stirling engine heat pipe condenser surfaces

The work done by Thermacore, Inc., Lancaster, Pennsylvania, for the Phase 1, 1992 SBIR National Aeronautics and Space Administration Contract, 'Insoluble Coatings for Stirling Engine Heat Pipe Condenser Surfaces' is described. The work was performed between January 1992 and July 1992. Stirling heat engines are being developed for electrical power generation use on manned and unmanned earth orbital and planetary missions. Dish Stirling solar systems and nuclear reactor Stirling systems are two of the most promising applications of the Stirling engine electrical power generation technology. The sources of thermal energy used to drive the Stirling engine typically are non-uniform in temperature and heat flux. Liquid metal heat pipe receivers are used as thermal transformers and isothermalizers to deliver the thermal energy at a uniform high temperature to the heat input section of the Stirling engine. The use of a heat pipe receiver greatly enhances system efficiency and potential life span. One issue that is raised during the design phase of heat pipe receivers is the potential solubility corrosion of the Stirling engine heat input section by the liquid metal working fluid. This Phase 1 effort initiated a program to evaluate and demonstrate coatings, applied to nickel based Stirling engine heater head materials, that are practically 'insoluble' in sodium, potassium, and NaK. This program initiated a study of nickel aluminide as a coating and developed and demonstrated a heat pipe test vehicle that can be used to test candidate materials and coatings. Nickel 200 and nickel aluminide coated Nickel 200 were tested for 1000 hours at 800 C at a condensation heat flux of 25 W/sq cm. Subsequent analyses of the samples showed no visible sign of solubility corrosion of either coated or uncoated samples. The analysis technique, photomicrographs at 200X, has a resolution of better than 2.5 microns (.0001 in). The results indicate that the heat pipe environment is not directly comparable to liquid metal pumped loop data, that nickel aluminide is still a leading candidate for solubility corrosion protection, and that longer duration tests are required to reach a definitive conclusion whether coatings are required at all. Should further testing be required, the test vehicle and analytical tools were developed.

Dussinger, Peter M.↗

Impact of Grain Size on Performance Degradation of TREAT LEU

We argue that radiation damage induced degradation of thermal conductivity does not set a lower limit on fuel grain sizes for the low enriched uranium fuel design of the Transient Reactor Test Facility (TREAT). Earlier work reports that smaller grains cause a larger degradation of thermal conductivity than larger grains constraining the smallest feasible size of fuel grains. This work assesses TREAT’s transient performance in the presence of radiation damage. The difference between the two studies is in treating damaged and fresh graphite as serial (this work) or parallel (previous) heat resistors. We use a multiphysics model of TREAT fuel grains to compute the reduction in transient capability measured by the total deposited energy as a function of irradiation dose. We find that radiation damage has a negligible effect on energy deposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Refabriation of HERA-HBU-2, LOC-HBU-1, and LOC-HBU-2

The development of advanced fuels for Light Water Reactors (LWRs) is essential for improved fuel performance and safety, including performance margins, uprates, and higher burnup. Because available commercial LWR fuels often cannot be used in research reactors, the refabrication of preirradiated fuel is necessary to enable follow-on testing of these fuels for characterizing fuel behavior during steady and transient irradiations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Flow Enhanced Electrochemical Sensor Performance in Complex Salt Systems

To advance MSR MC&A practices, Argonne National Laboratory developed and optimized robust flow enhanced electrochemical sensors (FEES) and multielectrode array voltammetry sensors (MAVS) which provide accurate near-real time measurements of actinides in molten salts. Flow-enhanced electrochemical sensors are installed directly into MSR flow conduits for inline salt chemistry measurement, while MAVS are deployed in quiescent salt conditions enabling online species concentration determination in stationary salt vessels. This report summarizes efforts to improve electrochemical sensor technological readiness through (1) sensor testing in complex, high concentration molten salt systems to demonstrate low uncertainty actinide measurements in MSR-representative solutions and (2) demonstrations of sensor performance in challenging real-world conditions in collaboration with partner MSR institutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

THOR-aLEU Report

The advanced low enriched uranium (aLEU)-THOR experiment vehicle is designed to irradiate fresh fuel rodlet specimens containing UO 2 (uranium dioxide) in the Temperature Heat-sink Overpower Response (THOR) capsule. The goal of the experiments are to compare the behaviors of standard UO 2 pellets to an aLEU concept in which molybdenum (Mo) foils separate UO 2 wafers increasing the thermal conductivity of the fuel. This experiment is designed to be conducted in the Transient Reactor Test facility (TREAT) using the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation system. The goal of these experiments is to assess the effect of thermally conductive inserts in UO 2 by measuring their net effect on radial thermal conductivity using transient nuclear heating and then to determine their power-to-melt threshold in transient overpower ramps. The objective of the aLEU-THOR fuels experiment is to determine the viability and study the performance of UO 2 fuel pellets employing newly upgraded TREAT capability. A novel transient measurement technique will be used with the THOR capsule, which has a solid heat sink to generate the transient temperature gradient for conductivity measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Correlations of the Steam Oxidation Rate Constant of BWR Alloy Zircaloy-2 at 800–1400 °C

Steam oxidation experiments were conducted at 800–1400 °C with boiling water reactor alloy Zircaloy-2 strip specimens. Sample weight gain measurements were performed on the oxidized specimens before and after the test and were compared to oxygen pickup calculations using the Cathcart–Pawel correlation. The results showed that Zircaloy-2 follows the parabolic law at temperatures above 1000 °C. At or below 1000 °C, the oxidation rate was very low when compared to Cathcart–Pawel correlation and can be represented by a cubic expression. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1000 °C and cubic rate constants are provided for temperatures at or below 1000 °C. Here, the weight gains calculated by our Arrhenius correlations are in excellent agreement with the measured sample weight gains at all test temperatures.

36 MATERIALS SCIENCE↗

Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data include, but are not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs. It is recognized that managing the data generated by large research and development projects presents a significant challenge for retaining data integrity and availability. American Society of Mechanical Engineers Standard NQA-1 (NQA-1) 2008/2009a provides appropriate requirements for this plan.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of a Chlorine Cathode and Anode Basket Assembly for Production of Uranium Chloride

A chlorine cathode has been developed for in situ chlorination of metals, oxides, and oxychlorides in molten chloride electrolytes that could be used to support synthesis of chloride fuel salts for molten salt reactors. The chlorine cathode is designed to electrochemically reduce chlorine gas to generate chloride ions that, when paired with a metal anode, chlorinate that metal as it is oxidized into the salt. The designed porous carbon electrode effectively distributes Cl 2 to the electrode surface and efficiently generates Clions in the molten chloride electrolyte. This report highlights recent improvements made to the electrode assembly, with a focus on operational control of the anode basket stability. Synthesis tests demonstrated the successful chlorination of uranium metal, resulting in 3.6 wt% uranium generated in the LiCl-KCl base salt in 45 minutes, performed in a bench-scale chlorination apparatus. Higher concentrations could be achieved by applying longer chlorination times or increasing the amount of uranium loaded in the anode basket. Overall, the chlorine cathode can be used with an appropriately designed anode to chlorinate uranium in situ for synthesis of molten salt reactor fuel salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Fabricating Fuel for the Versatile Test Reactor

A metal driver fuel has been proposed for the Versatile Test Reactor (VTR). About 30 years ago, the Experimental Breeder Reactor-II (EBR-II) was the last reactor in the U.S. to utilize a full core of metal driver fuel. While the necessary knowledge to make metal fuels is well preserved and is practiced for research activities today, re-establishing a production line to support the fuel needs of a 300 MWth reactor has unique technical and engineering challenges. These challenges are the focus of a multi-laboratory and private sector team that has been tasked with the responsibility to fabricate fuel for the VTR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Performance of Pulsed Thermal Tomography Imaging with Machine Learning-Based Classification of Defects in Additively Manufactured Structures

Additive manufacturing (AM) is an emerging method for cost-efficient fabrication of complex topology nuclear reactor parts from high-strength corrosion resistance alloys, such as stainless steel and Inconel. AM of metallic structures for nuclear energy applications is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Integrity of AM structures needs to be evaluated nondestructively because material flaws could lead to premature failures in high temperature nuclear reactor environment. Currently, there exist limited capabilities to evaluate actual AM structures non-destructively. Pulsed Thermal Tomography Imaging (PTT) provides a capability for non-destructive evaluation (NDE) of subsurface defects in arbitrary size structures. The PTT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PTT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PTT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PTT capabilities in detecting and visualizing microscopic defects in metallic specimens. The thermal tomography (TT) algorithm obtains depth reconstructions of spatial effusivity from the data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring of images with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images. CNN is trained on a database of TT images created for a set of simulated metallic structures with elliptical subsurface voids. Test of CNN performance demonstrate the ability to classify radii and angular orientation of subsurface defects in TT images. In addition, we have shown that CNN trained on elliptical defects is capable of classifying irregular-shaped defects obtained from scanning electron microscopy (SEM) of stainless steel sections printed with LPBF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

HERA M&S Exercise Problem Description Report

The Nuclear Energy Agency (NEA) Framework for Irradiation Experiments (FIDES) program includes the High burnup Experiments for Reactivity initiated Accident (HERA) Joint Experimental Program (JEEP). The HERA project is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity Initiated Accident (RIA) conditions. The HERA experiment plan includes analytical integral experiments using test specimens tailored to investigate specific conditions of relevance as well as prototypic integral experiments focused on irradiated fuel from prototypic origin. Modeling & simulation (M&S) is a key component of any experiment program, and the HERA JEEP is coordinating a M&S exercise. The purpose of this document is to provide problem descriptions to support the HERA M&S exercise based on fuel performance modeling. The HERA M&S exercise is expected to evolve into multiple efforts in outyears. This document may be revised and expanded to incorporate those evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Survey of baffle plate samples, specimen machining, and hydrogen measurements

This report outlines preliminary results on the survey and characterization of irradiated baffle plate, baffle former, and flux thimble tube specimens as part of the ongoing investigation into irradiation-induced embrittlement in austenitic stainless steels at ambient (room temperature) conditions. The specimens originated from commercial pressurized water reactor (PWR) components, covering displacement damage doses ranging from approximately 0.065 to 75 dpa. Initial scanning electron microscopy (SEM) surveys revealed that specimen surfaces exhibited fine machining marks and in-service-formed oxide layers on the side surfaces of the analyzed specimens. The oxide layers revealed specific features resembling localized "pitting-like" corrosion. An additional set of specimens representing in-service oxidation is being preserved for future microstructure analysis work. As believed, this will provide additional insights into long-term material degradation in PRWs. Specimen machining challenges emerged due to a complex failure of the electric discharge machining (EDM) system located in hot area. Despite partial restoration, issues persist with the EDM’s secondary power supply, necessitating the exploration of alternative EDM or computer numerical control (CNC) machining approaches to facilitate tensile specimen preparation. Currently, low speed saw cutting is in progress to prepare specimens for hydrogen measurements and general microstructure analysis. The near-term goals include completing tensile specimen machining for advanced mechanical testing and characterizing fracture mechanisms, and stress-corrosion cracking testing, ultimately aiming to identify and mitigate the ambient-condition intergranular cracking through targeted post-irradiation annealing strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography

The AGR-3/4 irradiation tests combined the third and fourth planned irradiation experiments in the US Department of Energy’s Advanced Gas Reactor (AGR) testing campaign of tri-structural isotropic (TRISO) fuel compacts. In this article, we present post-irradiation examination (PIE) using X-ray computed tomography (XCT) of two unirradiated and two irradiated compacts from the AGR-3/4 irradiation tests. The irradiated compacts studied (compact 7–1 and compact 12–4) represent the upper and lower limit of burnup within the AGR-3/4 irradiation experiment. This article presents a detailed quantitative analysis on the post-irradiation structure of TRISO fuel compacts. Various quantitative parameters including shape, size, and packing of kernels, and their spatial distribution, were utilized to gain insights into the structural changes caused by irradiation. The equivalent diameter and sphericity were found to increase and decrease, respectively, in irradiated compact 7–1 due to its higher burnup. Nearest neighbor distance between fuel kernels decreased after irradiation, suggesting irradiation-induced shrinkage of graphitic matrix. Furthermore, each compact in AGR-3/4 irradiation tests contained 20 designed-to-fail (DTF) fuel particles that were meant to act as a source of fission product release to the experiment test train. Furthermore, in the present work, all DTF fuel particles in the four compacts studied were identified, and it was found that they exhibited larger kernel swelling in compact 12–4 and smaller kernel swelling in compact 7–1, compared to the driver particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Design and performance of a variable gap system for thermal conductivity measurements of high temperature, corrosive, and reactive fluids

We report high-temperature fluids such as molten salts, liquid metals, and gasses are being proposed for many advanced energy systems including thermal energy storage devices, concentrating solar plants, and advanced nuclear reactor designs. However, the chemical behavior and thermophysical properties of many of these fluids have not been well characterized, which hinders the design, modeling, safety analysis, and deployment of these systems. Thermal conductivity is a property that is especially limited by existing measurement capabilities, which are subject to errors caused by convection, material interaction, radiative heat transfer, and instrument degradation. Therefore, there is a lack of standard, systematic measurement techniques for high-temperature, reactive, and corrosive fluids. In this work, the development of a variable gap thermal conductivity measurement system is detailed. The system is designed to measure the thermal conductivity of highly corrosive and reactive fluids, and survive operation between 100 °C and 800 °C. The effects of convection are minimized by limiting the thickness of the specimen to thin sizes (<0.3 mm). Corrections for radiative heat transfer were included in the working equations to consider specimens with varying optical properties. The design, construction, instrumentation, operating principles, and data analysis techniques are discussed in detail. The system was tested up to 500 °C using helium gas and molten KNO 3 -NaNO 3 to verify the measurement technique and determine the sources of error. At 300 and 400 °C KNO 3 -NaNO 3 , results showed maximum relative error of 6% when compared to results in the literature. The helium results were within 13% of those in the literature at 300 and 400 °C. Higher errors were observed at 500 °C for both fluids, and the sources of these errors are discussed.

42 ENGINEERING↗

Chemical vapor deposition of high T(sub c) superconducting films in a microgravity environment

Since the discovery of the YBaCuO bulk materials in 1987, Metalorganic Chemical Vapor Deposition (MOCVD) has been proposed for preparing HTSC high T(sub c) films. This technique is now capable of producing high-T(sub c) superconducting thin films comparable in quality to those prepared by any other methods. The MOCVD technique has demonstrated its superior advantage in making large area high quality HTSC thin films and will play a major role in the advance of device applications of HTSC thin films. The organometallic precursors used in the MOCVD preparation of HTSC oxide thin films are most frequently metal beta-diketonates. High T(sub c) superconductors are multi-component oxides which require more than one component source, with each source, containing one kind of precursor. Because the volatility and stability of the precursors are strongly dependent on temperature, system pressure, and carrier gas flow rate, it has been difficult to control the gas phase composition, and hence film stoichiometry. In order circumvent these problems we have built and tested a single source MOCVD reactor in which a specially designed vaporizer was employed. This vaporizer can be used to volatilize a stoichiometric mixture of diketonates of yttrium, barium and copper to produce a mixed vapor in a 1:2:3 ratio respectively of the organometellics. This is accomplished even though the three compounds have significantly different volatilities. We have developed a model which provides insight into the process of vaporizing mixed precursors to produce high quality thin films of Y1Ba2Cu3O7. It shows that under steady state conditions the mixed organometallic vapor must have a stoichiometric ratio of the individual organometallics identical to that in the solid mixture.

Levy, Moises↗