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The presentation covers work performed in FY23 on surveillance test article development. Surveillance test articles have been used of integrity assessment in light water reactor plants, this work seeks to develop a specimen that is passively loaded and could be used in advanced nuclear reactors for structural material health monitoring.
In the past several decades, numerous experiments have been performed worldwide to support reactor operations, measurements, design, and nuclear safety. Those experiments represent an extensive international investment in infrastructure, expertise, and cost, representing significantly valuable resources of data supporting past, current, and future research activities. Those valuable assets represent the basis for recording, development, and validation of our nuclear methods and integral nuclear data [1]. The loss of these experimental data, which has occurred all too much in the recent years, is tragic. The high cost to repeat many of these measurements can be prohibitive, if not impossible, to surmount. Two international projects were developed, and are under the direction of the Organisation for Co-operation and Development Nuclear Energy Agency (OECD NEA) to address the challenges of not just data preservation, but evaluation of the data to determine its merit for modern and future use. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was established to identify and verify comprehensive critical benchmark data sets; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data [2]. Similarly, the International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications [3]. Annually, contributors from around the world continue to collaborate in the evaluation and review of select benchmark experiments for preservation and dissemination. The extensively peer-reviewed integral benchmark data can then be utilized to support nuclear design and safety analysts to validate the analytical tools, methods, and data needed for next-generation reactor design, safety analysis requirements, and all other front- and back-end activities contributing to the overall nuclear fuel cycle where quality neutronics calculations are paramount.
Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.
Fiber optic and ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. These technologies, which have performed successfully in out-of-pile tests, have not been well qualified in a test reactor environment. The uncertainty of successful operation of instruments at this level of development has typically been a significant barrier to in-core deployment. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of these technologies to provide enhanced sensors for irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. The National Scientific User Facility (NSUF) is sponsoring a project to evaluate the performance of promising temperature sensors based on ultrasound and on fiber optics during an upcoming irradiation test in the Massachusetts Institute of Technology Research Reactor (MITR). Several types of thin-wire metallic waveguide ultrasonic thermometers will be tested along with fiber optic temperature sensors using regenerated Fiber-Bragg Grating technology. Additionally, samples of several new optical materials which are potentially much more radiation tolerant than commonly used materials, such as silica, will be included and analyzed post-irradiation. The irradiation is a collaborative effort between researchers from Idaho National Laboratory, the French Alternative Energies and Atomic Energy Commission, the University of Pittsburgh, the Massachusetts Institute of Technology, and AFO Research, Inc.
The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) was designed to perform non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and materials. NTREES can simulate the thermal hydraulic environment within the reactor of an NTR and has been leveraged to provide data on the thermochemical, thermomechanical and thermohydraulic performance of components and materials. Active upgrades to the NTREES facility include installing a DC power supply for DC powered joule heating, a supply gas chiller for a cryogenically cool gas supply, gas pre-heaters with sample holders for testing samples too small to be inductively heated by bathing them in hot gas, and a high temperature digital image correlation (DIC) capability for measuring high temperature strains real time. In addition to NTR testing, given the commonality of high enthalpy flow environments between the reactor of an NTP engine and the aerothermal heating of high-speed flight, the unique capabilities of NTREES to provide high temperature gas flows at high pressure without combustion or electrode spallation byproducts may lend itself to additional usage of the facility beyond the original intent.
Results of neutron dosimetry measurements in ATR for cycle 175B-1 are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.
This paper discusses some considerations for the proposed design and analysis of canisters used to ship reactor debris from the Fukushima Daiichi nuclear plant. Energy-limited events, such as cask drops and other dynamic impacts, are evaluated much differently than other structures using static analysis. Static analysis of structures makes sense for engineers because static analysis is intuitive and material typically behaves in a linear elastic manner. Dynamic analyses are typically not intuitive because of non-linear material behavior. This is why canister designs need to be evaluated by dynamic finite element analysis and/or laboratory drop tests.
SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. Although SAS4A/SASSYS-1 was developed to support the analysis of any liquid-metal-cooled nuclear reactor, it has primarily been utilized to design and analyze Sodium Fast Reactors (SFRs). As a result, most of the verification basis for SAS4A/SASSYS-1 utilizes sodium as a coolant and geometry descriptions that are typical of a SFR facility. In this work, new verification problems are defined to extend the applicability of the SAS4A/SASSYS-1 verification basis into Lead Fast Reactor (LFR) design space. A review of the existing SAS4A/SASSYS-1 SFR verification test suite is performed to determine the number of test cases that are SFR specific. SFR specific test cases are replicated using lead as a coolant and system layouts that are representative of LFRs. This report provides definitions and reference solutions for seven new test cases that scope lead-specific features and capabilities of the software. For a verification test to be considered acceptable, comparisons of analytical solutions to SAS4A/SASSYS-1 predictions must produce negligible or justifiable errors. All LFR specific test cases are found to be acceptable.
This University Engagement project challenged engineering students at universities, that do not have Bachelor degree programs in nuclear engineering but do have research reactors and some nuclear engineering coursework, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative (NGSI), Human Capital Development (HCD): University Engagement Program. This program exposed university students with Mechanical Engineering majors and Nuclear Engineering minors to the concepts of international nuclear safeguards. In FY22, three teams at the University of Rhode Island and two teams at the University of Texas - Austin participated in researching, designing, building, and testing projects to support international nuclear safeguards measurements or verification. The projects involved engaging in activities at the university’s research reactors. All the projects engaged students with prototyping a design and/or tool for application at the Universities’ reactor. At the end of the course, most of the students expressed the experience was positive and they learned more about international nuclear safeguards and applying requirements than they had previously encountered. This school year the projects were further complicated by the COVID-19 pandemic. Both universities had limited in classes on campus, still relying on Zoom classes, and limited direct student/professor interactions. Furthermore, Los Alamos National Laboratory (LANL) greatly restricted travel, therefore making it impossible to visit the students at the end of the semester for the review of their design projects. The final design and review meeting for the projects happened via meetings over the internet. Additionally, the teams did build and test some prototypes but could only do so in a limited capacity.
The Utah State University (USU) Advanced Test Reactor (ATR) National Scientific User Facility experiment is a materials test consisting of specimens of hafnium aluminide intermetallic and composite. Specimens were contained in sealed, drop-in capsule pairs that were irradiated in the ATR B2 position for up to four reactor cycles (149A, 149B, 150 B and 151A). The experiment configuration was modeled using ABAQUS CAE Version 6.14-2 for construction of the finite element model and ABAQUS Standard to calculate specimen temperatures under as-run irradiation conditions. This ECAR report was formalized as requested by the NSUF project manager (Attachment A). A thermal analysis was performed to determine the irradiation temperatures of each test specimen for up to four irradiation cycles (149A, 149B, 150 B and 151A) that comprised the irradiation campaign. Computed minimum, average and maximum specimen temperatures are given for each of the 132 specimens for an average set of heat rates provided by the physics analysis. In addition, bounding temperatures are given for each specimen using the maximum heat rates from the physics analysis. Specimens in Capsules C 1, D 1 and B 2 experienced the lowest temperatures, followed by those in Capsules B 1, C 2 and D-2, and the highest temperatures were in specimens located in Capsules A-1 and A-2. Temperatures of the specimens for all cycles were between 325.2K and 424.7K. The highest average specimen temperature of 422.9K occurred in the intermetallic rod specimens in Capsule A-1 during Cycle 149B. The highest bounding temperature in the specimens was 424.7K. This information will be used in conjunction with PIE measurements to elucidate the thermophysical mechanisms occurring during, and as a result of, irradiation in the ATR.
The microstructural evaluation and characterization of 316 stainless steel samples that were tested in molten Li 2 BeF 4 (FLiBe) salt were investigated in this study for evaluating its performance in high-temperature molten fluoride salts. Recently, 316 stainless steel and FLiBe salt are being actively considered as the main structural alloy and primary coolant of fluoride salt-cooled high-temperature reactor (FHR), a leading nuclear reactor concept for the next generation nuclear plants (NGNP). In support of the materials development for the FHR, high-temperature corrosion tests of 316 stainless steel in molten FLiBe salt at 700°C have been conducted in both bare graphite crucibles and 316 stainless steel-lined crucibles in an inert atmosphere for up to 3000 hours. The microstructure of the tested samples was comprehensively characterized using scanning electron microscopy (SEM) in conjunction with energy dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD), and scanning transmission electron microscopy (STEM) with EDS. In addition to the noticeable intergranular corrosion attack on surface, the corrosion in terms of the Cr depletion along high angle grain boundaries (15-180º) extended to 22µm in depth after 3000-hour exposure to molten FLiBe salt in graphite crucible. The coherent Σ3 grain boundary appeared high resistance to the Cr depletion. The substantial Cr depletion from the near-to-surface layer induced phase transformation from γ-martensite to α-ferrite phase (FeNi x ) during corrosion at 700ºC. Furthermore, the presence of graphite in the molten salt doubled the corrosion attack depth and led to the formation of round Mo2C, hexagonal Cr 7 C 3 and needle-like Al 4 C 3 phase within the alloy as deep as 50 µm after 3000-hour corrosion testing. Based on the microstructural analysis, the corrosion mechanisms of 316 stainless steel in molten FLiBe salt in different corrosion crucibles were illuminated through schematic diagrams. Additionally, a thermal diffusion controlled corrosion model was developed and validated by experimental data for predicting the long-term corrosion attack depth.
This poster has been developed for display during the 2023 Intern Poster Session. The poster is apart of the Nuclear Research and Experiments category. The focus of this poster is on the modifications made to the design of an integral effect test (IET) facility currently being designed for a new Small Modular Reactor (SMR) design. My contributions made to the IET facility are laid out along with an explanation on some of the aspects of designing a new nuclear reactor. This poster covers the ideas and importance of scaling and IET facilities with the context of the standards set by the NRC. There are also graphics of some other examples of IET facilities developed for other reactor designs. My contributions include designing the piping and instrumentation layout of the IET facility using CAD software. This is a key step in the process of designing this IET facility and needs to be designed to avoid excessive head loss.
The Department of Energy Office of Nuclear Energy (DOE-NE) vision is to “Advance nuclear energy science and technology to meet U.S. energy, environmental, and economic needs.” The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy.
During the Nuclear Fuel Manufacturing process a number of by-products and residues are generated from various sources. In recent years, Westinghouse has characterised and processed around 50,000 drums of residue comprising of numerous different types generated over 60 years from the manufacture of numerous types of fuel for greater than 50 reactors. These include, AGR, Magnox (uranium metal fuel), PWR, fast reactors and various test reactors. Different organisations have different definitions and names for each of these categories but essentially they fall into one of the following categories: Clean Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Dirty Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Gd Contaminated Scrap (U{sub 3}O{sub 8}, UO{sub 2}, powders, pellets), Ash, Sludge, Combustible / soft residues (e.g. ventilation filters, gloves, plastic bags), Contaminated metals (including contaminated scrap zirconium tubes), Materials arising from decommissioning. Routes have been developed across the industry for a number of the categories listed above, however in numerous examples, routes have not been established and materials have been stored creating a significant legacy. Westinghouse understands the issues regarding legacy uranic material and the financial and regulatory impact of storage. Interim storage of materials may seem like a low cost option in the short term; however ongoing costs for facilities, manpower and equipment, without a permanent and final solution can result in significant lifetime material management costs. Recovery of uranium and sending materials for safe disposal of enriched uranium residues is not easy because of the criticality risk both during processing and at a disposal site. Therefore quantities of materials than can be sent via permitted disposal routes are very limited The development of regulations often means that finding solutions is becoming more difficult. It is believed that timely resolution is ultimately the most cost effective solution. In addition, the intrinsic value of recycling this material is also high and in some cases can mitigate the cost of processing. (authors)
The effect of irradiation on materials properties is an important field of study for materials usage in both fission and fusion systems for energy production. Neutron flux and energy spectrums are well understood; however, the irradiation temperature can be more difficult to determine. Since the early 1960s, Silicon Carbide (SiC) has been used as a passive post-irradiation temperature monitor because the irradiation defects anneal out above the irradiation temperature. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process. There are many properties that may be measured including electrical resistivity, bulk density, dimensions, thermal diffusivity, or lattice spacing. Electrical resistivity is accepted as a robust measurement technique; however, such method is time-consuming since the steps involved must be performed in a serial manner. The use of thermal expansion from continuous dilatometry is an automated process requiring minimal setup time. As part of a Nuclear Science User Facilities (NSUF) project, low dose silicon carbide monitors were irradiated in the BR2 material test reactor at SCK, Belgium, and were then evaluated at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This paper will discuss the evaluation processes available at HTTL and the performance of the low dose silicon carbide temperature monitors.
Research on molten salt reactor (MSR) technology based on fluoride salts as a convenient way of higher actinides incineration and/or utilization of thorium fuel has been performed at UJV Rez and later Research Centre Rez in Czech Republic since 1999. Necessary steps towards to the deployment and possible commercialization were examined in past investigations. Neutronic testing is an inseparable part of the development and was at first carried out in a series of experiments with LiF-NaF salt (60 mol% LiF, 40 mol% NaF) subjected to a neutron field shaped by different moderators within the framework of the EROS experiments. Later, neutron spectrum was measured in this salt and significant variations in calculation to experiment C/E comparison were discovered. Further tests were performed with Flibe (66 mol% LiF, 33 mol% BeF{sub 2}) reactor coolant salt containing enriched lithium (99.994 mol% Li-7), which was provided by Oak Ridge National Laboratory (ORNL). This Flibe salt originally comes from the Molten Salt Reactor Experiment (MSRE) and was supplied to the Czech Republic within the framework of a collaborative agreement on MSR and fluoride-salt-cooled high-temperature reactor (FHR) research and development (R and D) concluded between the Czech Ministry of Industry and Trade and US Department of Energy in 2012. All experiments in this work were carried out at the LR-0 reactor at Research Centre Rez. The LR-0 is a pool-type zero power light water reactor with large and flexible experimental capabilities due to its design and fuel construction.
Plate-fin heat exchangers for nuclear reactor Brayton cycle
Uncertainty and sensitivity analysis is an indispensable element of any substantial attempt in reactor simulation validation. The quantification of uncertainties in nuclear engineering has grown more important and the IAEA Coordinated Research Program (CRP) on High-Temperature Gas Cooled Reactor (HTGR) initiated in 2012 aims to investigate the various uncertainty quantification methodologies for this type of reactors. The first phase of the CRP is dedicated to the estimation of cell and lattice model uncertainties due to the neutron cross sections co-variances. Phase II is oriented towards the investigation of propagated uncertainties from the lattice to the coupled neutronics/thermal hydraulics core calculations. Nominal results for the prismatic single block (Ex.I-2a) and super cell models (Ex.I-2c) have been obtained using the SCALE 6.1.3 two-dimensional lattice code NEWT coupled to the TRITON sequence for cross section generation. In this work, the TRITON/NEWT-flux-weighted cross sections obtained for Ex.I-2a and various models of Ex.I-2c is utilized to perform a sensitivity analysis of the MHTGR-350 core power densities and eigenvalues. The core solutions are obtained with the INL coupled code PHISICS/RELAP5-3D, utilizing a fixed-temperature feedback for Ex. II-1a.. It is observed that the core power density does not vary significantly in shape, but the magnitude of these variations increases as the moderator-to-fuel ratio increases in the super cell lattice models.