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DOE OSTI2017

Assessment of Sensor Technologies for Advanced Reactors

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope for fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.

Korsah, Kofi↗
DOE OSTI2016

Embedded Sensors and Controls to Improve Component Performance and Reliability -- Loop-scale Testbed Design Report

Embedded instrumentation and control systems that can operate in extreme environments are challenging to design and operate. Extreme environments limit the options for sensors and actuators and degrade their performance. Because sensors and actuators are necessary for feedback control, these limitations mean that designing embedded instrumentation and control systems for the challenging environments of nuclear reactors requires advanced technical solutions that are not available commercially. This report details the development of testbed that will be used for cross-cutting embedded instrumentation and control research for nuclear power applications. This research is funded by the Department of Energy's Nuclear Energy Enabling Technology program's Advanced Sensors and Instrumentation topic. The design goal of the loop-scale testbed is to build a low temperature pump that utilizes magnetic bearing that will be incorporated into a water loop to test control system performance and self-sensing techniques. Specifically, this testbed will be used to analyze control system performance in response to nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics and gyroscopic effects, and impeller disturbances. This testbed will also be used to characterize the performance losses when using self-sensing position measurement techniques. Active magnetic bearings are a technology that can reduce failures and maintenance costs in nuclear power plants. They are particularly relevant to liquid salt reactors that operate at high temperatures (700 C). Pumps used in the extreme environment of liquid salt reactors provide many engineering challenges that can be overcome with magnetic bearings and their associated embedded instrumentation and control. This report will give details of the mechanical design and electromagnetic design of the loop-scale embedded instrumentation and control testbed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Digital-Twin-Enabling Technologies for Online Condition Monitoring of Nuclear Power Plant Components

Online condition monitoring is an area of active research that may enable improved scheduling, maintenance, and safety of nuclear power plant components, reducing unnecessary derates while simultaneously improving operational capacity. Digital twins are one avenue to conduct online condition monitoring and are currently being explored by national laboratories and universities alike. Digital twins for online condition monitoring are, in essence, state concurrent models that emulate a physical process which predicts a parameter and compares it against a measured value. However, digital twins may also provide additional insights by combining and interpreting various sources of information. These insights may be used for preventative maintenance scheduling optimization or early fault detection and are projected to be valuable for meeting requirements under 10 CFR 50.55a. However, digital twin technologies are still under significant development; quantifying model uncertainties, improving unique fault identification, and multimodal sensor fusion are some areas under investigation. Therefore, in this work, we discuss and review the various enabling technologies, in the form of advanced sensors, instrumentation, and modelling methods, that may be used to implement and enhance digital twins for online condition monitoring. A potential use case for pump-motors is presented to demonstrate how these various pieces of enabling digital twin technologies may integrated together for online condition monitoring. Challenges and opportunities associated with the pump-motor digital twin enabling technologies are also identified and discussed.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Design Considerations for Secure Wireless Sensor Communication Systems in Harsh Electromagnetic Environments of Nuclear Reactor Facilities

Wireless sensors can potentially play a significant role in safety, efficiency, and reliability of the instrumentation and control process in current and next generation nuclear power reactors. While conventional narrowband wireless sensors have shown a certain level of success in some nuclear power plants (NPPs), the radio frequency (RF) propagation challenges posed by the heavy metallic and cluttered environment of NPPs has prevented their widespread use in such operations. Furthermore, these challenges include RF wave propagation in harsh (reflective, absorptive, cluttered) environments, data security issues, and RF interference to and from other devices in the vicinity of a nuclear reactor core. In this paper, first we address how ultrawideband (UWB) RF technology can complement the narrowband (i.e., WiFi) solutions that have been used in some NPPs by providing an alternative solution in addressing the signal propagation issues in such electromagnetically harsh environments. Second, we discuss and present the UWB software simulation results on multipath harsh environments, and then address the data security issues. In the final sections of the paper, we present the experimental results of using UWB signaling in a representative harsh environment conducted at the Massachusetts Institute of Technology research reactor site. We plan to develop the UWB communications hardware based on the results of this paper and report on its performance in the field with emphasis on the security aspects of the system in a subsequent paper.

42 ENGINEERING↗
DOE OSTI2017

Enhanced electrical penetration components meet the high safety standards required for China's HTR-PM Shidaowan and generation mPower small modular reactors - 300

Enhanced glass-to-metal feedthroughs will be used to carry key instrumentation and control (I and C) signals in China's high-temperature, gas-cooled pebble-bed modular reactor (HTR-PM) at Shidaowan, and in Small Modular Reactor (SMR) prototyping. Chinergy Co., Ltd and Jiamusi Electric Machine Co., Ltd will use the feedthroughs, also called electrical penetration assemblies (EPAs), at the Shidaowan twin-reactor high temperature reactor (HTR). These feedthroughs are manufactured using specialized processes that make them far more robust than counterparts made with organic or ceramic materials. Improved EPAs manufactured by SCHOTT are the only feedthrough solution that can be used in the primary loop of an HTR because organic seals cannot withstand the high temperatures and pressures of this environment. Glass-to-metal seals (GTMS) are made using inorganic, non-aging materials which enable a virtually unlimited lifespan in high-temperature, high-radiation and high-pressure applications. This I and C technology package is being enhanced by the development and testing of new mineral-insulated (MI) cable connectors and EPAs made of crystallized glass-ceramic combined with superalloys. The EPAs have a maintenance-free lifespan of over 60 years, preventing any outage time related to these components and subsequently reducing costs associated with shutdowns. They are far more robust than organic seals and provide increased safety at a lower lifetime cost. Resistance against high pressure and high temperature are especially ideal qualities that make EPAs a valuable choice for environments with high standards for fire protection, such as nuclear plants, subsea oil drilling, or any other underwater application. Due to their significant offering of safety-relevant qualities, the EPAs have also been chosen by BWXT for implementation in its mPower SMR. Small modular reactor requirements are complex and very challenging but are met by the new EPA components, which will help form a strong I and C safety chain. Unlike ordinary compression glass-to-metal seals, GTMS with crystallized glass-ceramic have the ability to withstand temperatures of over 350 deg. C / 662 deg. F and operating pressures of 3500 bar. Furthermore, their lifetime is qualified for up to 60 years, including pressure and radiation requirements in accordance with the latest IEC/IEEE standards. Higher-specification EPAs and MI cable connectors have been designed into severe accident I and C instruments, including advanced hydrogen and oxygen sensors that exceed the challenging requirements of both HTR and SMR applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development, Testing and Validation of a Waste Assay System for the Measurement and Characterisation of Active Spent Fuel Element Debris From UK Magnox Reactors - 12533

This paper describes the development, testing and validation of a waste measurement instrument for characterising active remote handled radioactive waste arising from the operation of Magnox reactors in the United Kingdom. Following operation in UK Magnox gas cooled reactors and a subsequent period of cooling, parts of the magnesium-aluminium alloy cladding were removed from spent fuel and the uranium fuel rods with the remaining cladding were removed to Sellafield for treatment. The resultant Magnox based spent fuel element debris (FED), which constitutes active intermediate level waste (ILW) has been stored in concrete vaults at the reactor sites. As part of the decommissioning of the FED vaults the FED must be removed, measured and characterised and placed in intermediate storage containers. The present system was developed for use at the Trawsfynydd nuclear power station (NPS), which is in the decommissioning phase, but the approach is potentially applicable to FED characterisation at all of the Magnox reactors. The measurement system consists of a heavily shielded and collimated high purity Germanium (HPGe) detector with electromechanical cooling and a high count-rate preamplifier and digital multichannel pulse height analyser. The HPGe based detector system is controlled by a software code, which stores the measurement result and allows a comprehensive analysis of the measured FED data. Fuel element debris is removed from the vault and placed on a tray to a uniform depth of typically 10 cm for measurement. The tray is positioned approximately 1.2 meters above the detector which views the FED through a tungsten collimator with an inverted pyramid shape. At other Magnox sites the positions may be reversed with the shielded and collimated HPGe detector located above the tray on which the FED is measured. A comprehensive Monte Carlo modelling and analysis of the measurement process has been performed in order to optimise the measurement geometry and eliminate interferences from radioactive sources and FED in the immediate vicinity of the measurement position. The detector system has been calibrated and high activity radioactive sources of Cs-137, Co-60 and Na-22 have been used to validate the measurement process. The data acquisition and analysis software code has been tested and validated in keeping with the software quality assurance requirements of both ISO:9001-2008 - TICK-IT in the UK and NQA-1. The measurement and analysis system has been comprehensively tested with high activity sources, is flexible and may be applicable to a wide range of remote handled radioactive waste measurement applications. It is due to be installed at Trawsfynydd NPS later this year. This paper describes the Waste Tray Assay System (WTAS) that has been developed for the measurement of Magnox FED waste. The WTAS has been tested with a range of radioactive sources and its operation has been simulated with benchmarked MCNP Monte Carlo calculations. The measurement software has been validated as has the operation of the system for a range of strong radioactive sources. A system based on the design is due for installation and operation in 2012. The system has application to the measurement of Magnox Fuel Element Debris (FED) waste at other Magnox reactor sites. The major design objective of the WTAS that has been achieved is the ability of the assay system to determine the content of Cs-137, and in turn to enable the fissile burden to be assessed using a radionuclide fingerprint, in the presence of higher and highly variable quantities of Co-60, typically from nimonic springs. The approach can be used in other Magnox FED waste configurations where the detector is located above the FED waste sorting tray and where the collimation is fixed below the detector and at a distance above the tray. In this case, which has also been investigated, there are different shielding problems and mechanical support issues. The extensive use of MCNP Monte Carlo modelling to simulate the geometry of the sorting cell and the distribution of radioactive sources has helped to ensure that all of the detector shielding requirements are addressed and suitable Cs-137 and Co-60 discrimination can be achieved. The WTAS in its present form or in other configurations has relevance to the measurement of other active ILW and highly active RH waste. Examples include high activity RH LLW and RH TRU (Transuranic) waste as defined in the United States arising from both commercial nuclear and Department of Energy (DOE) operations. The analysis is able to analyse a range of radionuclides beyong those expected in the Magnox FED cases. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Design Optimization for Printed Melt Wire Arrays Encapsulation

As part of the Nuclear Energy Enabling Technology (NEET) Advanced Sensor and Instrumentation (ASI) Program, Idaho National Laboratory (INL) has recently established in-house capabilities to fabricate and test new advanced-manufactured sensors for measuring peak irradiation temperature within a nuclear test reactor. Although methods of real-time temperature monitoring, such as thermocouples, may be used, the complexity of feedthroughs and attachments to collect real-time measurements greatly increases the cost of the experiment. Instead, passive monitoring techniques may be used for peak- temperature measurement that exploit the melting point of well-characterized materials (standard melt wires) to infer peak reactor temperatures. However, limited available space for instrumentation during experiments introduces an additional challenge. To accommodate this, INL has expanded its temperature- detection instrumentation capabilities to include advanced manufactured (AM) melt wires for peak irradiation temperature measurements. These melt wires can determine peak temperatures while also accommodate space limitations in irradiation experiments. In an effort to improve performance reliability of AM meltwire capabilities, a process was developed and tested to identify the significance of entrapping a high purity inert atmosphere within the packaging of printed melt wire arrays. The materials used in this study were aluminum, zinc, and tin encapsulated in high purity helium within a stainless steel (SS) 316 container. Tin, with a low melting point of approximately 230°C, Zn with a mid-melting point of approximately 420°C, and Al with a high melting point of approximately 660°C. This report describes the design, fabrication process, furnace testing and X-ray Computed Tomography (XCT) evaluation. Results show a successful outcome in creating an inert gas encapsulation and high-resolution evaluation methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

In-Pile Instrumentation Multi- Parameter System Utilizing Photonic Fibers and Nanovision

An advanced in-pile multi-parameter reactor monitoring system is being proposed in this funding opportunity. The proposed effort brings cutting edge, high fidelity optical measurement systems into the reactor environment in an unprecedented fashion, including in-core, in-cladding and in-fuel pellet itself. Unlike instrumented leads, the proposed system provides a unique solution to a multi-parameter monitoring need in core while being minimally intrusive in the reactor core. Detector designs proposed herein can monitor fuel compression and expansion in both the radial and axial dimensions as well as monitor linear power profiles and fission rates during the operation of the reactor. In addition to pressure, stress, strain, compression, neutron flux, neutron spectra, and temperature can be observed inside the fuel bundle and fuel rod using the proposed system. The proposed research aims at developing radiation-hard, harsh-environment multi-parameter systems for insertion into the reactor environment. The proposed research holds the potential to drastically increase the fidelity and precision of in-core instrumentation with little or no impact in the neutron economy in the reactor environment while providing a measurement system capable of operation for entire operating cycles.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Automated Work Packages Prototype: Initial Design, Development, and Evaluation. Light Water Reactor Sustainability Program

The goal of the Automated Work Packages (AWP) project is to demonstrate how to enhance work quality, cost management, and nuclear safety through the use of advanced technology. The work described in this report is part of the digital architecture for a highly automated plant project of the technical program plan for advanced instrumentation, information, and control (II&C) systems technologies. This report addresses the DOE Milestone M2LW-15IN0603112: Describe the outcomes of field evaluations/demonstrations of the AWP prototype system and plant surveillance and communication framework requirements at host utilities. A brief background to the need for AWP research is provided, then two human factors field evaluation studies are described. These studies focus on the user experience of conducting a task (in this case a preventive maintenance and a surveillance test) while using an AWP system. The remaining part of the report describes an II&C effort to provide real time status updates to the technician by wireless transfer of equipment indications and a dynamic user interface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Neutrino-detector design for safeguarding small modular reactors

Nuclear reactors have long been a favored source for antineutrino measurements for estimates of power and burnup. With appropriate detector parameters and background rejection, an estimate of the reactor power can be derived from the measured antineutrino event rate. Antineutrino detectors are potentially attractive as a safeguards technology that can monitor reactor operations and thermal power from a distance. Advanced reactors have diverse features that may present challenges for current safeguards methods. By comparison, neutrino detectors offer complementary features, including a remote, continuous, unattended, and near-real-time monitoring capability, that may make them useful for safeguarding certain classes of advanced reactors. This study investigates the minimum depth and size of an antineutrino detector for a small modular reactor to meet safeguards needs for advanced reactors. Furthermore, extrapolating performance from several prior reactor antineutrino experiments, this study uses an analytical approach to develop a possible design for a remote antineutrino-based monitoring device.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

New Features of TRIPOLI-4{sup R} Version 10

TRIPOLI-4{sup R} is a three-dimensional and continuous-energy Monte-Carlo transport code developed by the 'Service d'Etudes des Reacteurs et de Mathematiques Appliquees' (SERMA), at the CEA Paris-Saclay Center. It is devoted to shielding, reactor physics, criticality safety and nuclear instrumentation. TRIPOLI-4{sup R} has been developed from the mid 1990's in C++. A new version of the code is typically released every two years. In this paper, we present the new features of TRIPOLI- 4{sup R} version 10, to be soon released to NEA and RSICC.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

First of all, we will present the main principle of the method. A piezoelectric transducer, driven by a pulse generator, generates the acoustic waves in a cavity that may be the fuel rod or a chamber connected to an instrumented rod. The composition determination consists in measuring the time of flight of the acoustic signal emitted. The pressure can be estimated by a calibration process, above the measurement of the amplitude of the signal. Two projects will then be detailed. The first project consists in the development of advanced instrumentation for in-pile experiments in Material Testing Reactor. It constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. This acoustic method was tested with success during a first experiment called REMORA 3, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). As a first step of the development program, we performed in-pile tests on the most sensitive component, i.e., the piezoelectric transducer. For this purpose, the active part of this sensor has been qualified on gamma and neutron radiations and at high temperature. Various industrial piezo-ceramics were exposed to a high activity Cobalt source for few days. The cumulated dose was ranged from 50 kGy up to 2 MGy. Next, these devices were placed inside a Material Test Reactor to investigate their reliability towards neutron fluence. The final fluence after 150 days of irradiation was up to 1.6.10{sup 21}n/cm{sup 2} (for thermal neutron). Irreversible variations have been measured. Next, a specific sensor has been implemented on an instrumented fuel rod and tested in the frame of a REMORA 3 Irradiation test. It was the first experiment under high mixed, temperature neutron and gamma flux. A first irradiation phase took place in March 2010 in the OSIRIS reactor and in November 2010 for the second step of the irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas pressure and composition measurement by an acoustic sensor was conducted successfully between 2008 and 2010 on 5 high burn-up MOX fuel rods and 2 very high burn-up UO{sub 2} fuel rods in LECA Facility at Cadarache Centre. An improvement of this sensor has been proposed, allowing us to divide by two the uncertainty on the pressure measurement. In the case of hot-cell measurements, viscous liquid can be used to couple the sensor with the rod. For gas content with a pressure exceeding 15 bars and a 10% Xe/Kr ratio, such coupling may reduce relative acoustic method accuracy by ±7% for pressure measurement result and ±0.25 % for the assessment of gas composition. These results make it possible to demonstrate the feasibility of the technique on LWR fuel rods. The transducer and the associated methodology are now operational for non-destructive measurements in hot lab facilities and allow characterising the fission gas without puncturing the fuel rods. Up to now, any other non-destructive method can be proposed. A next step will be the development of an industrial application in a fuel storage pool in order to perform a large number of measurements on a fuel assembly in nuclear plants.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

The high intensity diffractometer for residual stress analysis (HIDRA), a third generation residual stress mapping neutron diffractometer at the high flux isotope reactor

This paper describes the hardware and software upgrades, operation, and performance of the high intensity diffractometer for residual stress analysis (HIDRA) instrument, a residual stress mapping neutron diffractometer located at the High Flux Isotope Reactor at Oak Ridge National Laboratory in Oak Ridge Tennessee, USA. Following a major upgrade in 2018, the new instrument has a single 3 He multiwire 2D 30 × 30 cm 2 position sensitive detector, yielding a field of view of 17° 2θ. Further, the increase in the field of view (from 4° 2θ) from the previous model instrument has contributed to the tremendous improvement in the out of plane solid angle such that the 3D count rate could be obtained easily. Accordingly, the hardware, software, Data Acquisition System (DAS), and so on have also been updated. Finally, all these enhanced features of HIDRA have been ably demonstrated by conducting multi directional diffraction measurements in the quenched 750-T74 aluminum, and the evolved and improved strain/stress mappings are presented.

3D printing↗
DOE OSTI2014

DEVELOPMENT OF OPERATIONAL CONCEPTS FOR ADVANCED SMRs: THE ROLE OF COGNITIVE SYSTEMS ENGINEERING

Advanced small modular reactors (AdvSMRs) will use advanced digital instrumentation and control systems, and make greater use of automation. These advances not only pose technical and operational challenges, but will inevitably have an effect on the operating and maintenance (O&M) cost of new plants. However, there is much uncertainty about the impact of AdvSMR designs on operational and human factors considerations, such as workload, situation awareness, human reliability, staffing levels, and the appropriate allocation of functions between the crew and various automated plant systems. Existing human factors and systems engineering design standards and methodologies are not current in terms of human interaction requirements for dynamic automated systems and are no longer suitable for the analysis of evolving operational concepts. New models and guidance for operational concepts for complex socio-technical systems need to adopt a state-of-the-art approach such as Cognitive Systems Engineering (CSE) that gives due consideration to the role of personnel. This approach we report on helps to identify and evaluate human challenges related to non-traditional concepts of operations. A framework - defining operational strategies was developed based on the operational analysis of Argonne National Laboratory’s Experimental Breeder Reactor-II (EBR-II), a small (20MWe) sodium-cooled reactor that was successfully operated for thirty years. Insights from the application of the systematic application of the methodology and its utility are reviewed and arguments for the formal adoption of CSE as a value-added part of the Systems Engineering process are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Data Collection Methods for Validation of Advanced Multi-Resolution Fast Reactor Simulations

In pool-type Sodium Fast Reactors (SFR) the regions most susceptible to thermal striping are the upper instrumentation structure (UIS) and the intermediate heat exchanger (IHX). This project experimentally and computationally (CFD) investigated the thermal mixing in the region exiting the reactor core to the UIS. The thermal mixing phenomenon was simulated using two vertical jets at different velocities and temperatures as prototypic of two adjacent channels out of the core. Thermal jet mixing of anticipated flows at different temperatures and velocities were investigated. Velocity profiles are measured throughout the flow region using Ultrasonic Doppler Velocimetry (UDV), and temperatures along the geometric centerline between the jets were recorded using a thermocouple array. CFD simulations, using COMSOL, were used to initially understand the flow, then to design the experimental apparatus and finally to compare simulation results and measurements characterizing the flows. The experimental results and CFD simulations show that the flow field is characterized into three regions with respective transitions, namely, convective mixing, (flow direction) transitional, and post-mixing. Both experiments and CFD simulations support this observation. For the anticipated SFR conditions the flow is momentum dominated and thus thermal mixing is limited due to the short flow length associated from the exit of the core to the bottom of the UIS. This means that there will be thermal striping at any surface where poorly mixed streams impinge; rather unless lateral mixing is ‘actively promoted out of the core, thermal striping will prevail. Furthermore we note that CFD can be considered a ‘separate effects (computational) test’ and is recommended as part of any integral analysis. To this effect, poorly mixed streams then have potential impact on the rest of the SFR design and scaling, especially placement of internal components, such as the IHX that may see poorly mixed streams. Finally, due to lack or infrastructural support for carrying out sodium experiments, only water experiments and CFD studies were realized in, an otherwise sodium approved facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Thermal deflection in neutron scattering sample environments at Oak Ridge National Laboratory

The neutron sources at Oak Ridge National Laboratory use a wide suite of sample environment equipment to deliver extreme conditions for a number of experiments. Much of this instrumentation focuses on extremes of temperature, such as cryostats, closed-cycle refrigerators in both low and high temperature configurations, and radiant heating furnaces. When the temperature is controlled across a large range, thermal deflection effects can notably move the sample and affect its alignment in the beam. Here, we combine these sample environments with neutron imaging and machine vision to determine the motion of a representative sample with respect to the neutron beam. We find vertical sample displacement on the order of 1–2 mm and horizontal displacement that varies from near-negligible to 1.2 mm. While these deflections are not relevant for some of the beamlines at the first target station at the spallation neutron source and the high flux isotope reactor, they will become critical for upcoming instrumentation at the second target station, as well as any instruments targeting sub-mm samples, as neutron sources and optics evolve to smaller and more focused beams. We discuss mitigation protocols and potential modifications to the environment to minimize the effect of misalignment due to thermal deflection.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Risk-informed Graded Approach for Reliability and Performance Assessment of Sensor and Instrumentation Systems within Advanced Condition Monitoring Technologies

Advanced condition monitoring (ACM) technologies, such as digital twins, are innovative strategies designed to provide real-time health insights, including the remaining useful life of components. The primary goal of ACM is to predict and alert operators to potential functional failures before they occur. ACM systems achieve this by integrating predictive models with various sensor instrumentation, analog-to-digital converters, data warehouses, and data pre-processors. These sensor and instrumentation systems (SIS) are essential for forming a comprehensive understanding of component conditions and ensuring the predictive success of ACM programs. Introducing new technologies like ACM involves varying degrees of risk that can impact plant reliability. Therefore, risk mitigation should be commensurate with the performance and reliability of the developed technology, following a risk-informed graded approach (RIGA). Establishing a RIGA process requires a clear understanding of the hazards and reliability of all subsystems, including their interdependencies and potential impacts on the overall system. Given the critical role of SIS in ACM, this work reviews hazard identification and reliability quantification methods for SIS. It also considers these methods' implications when developing a RIGA process for ACM.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Uncertainty Propagation from Experiment Measurements to Modeling Approaches: A Case for SMR Steam Entrainment Testing

To license new and advanced reactor designs, regulators must be convinced that their unique safety cases—relative to existing large scale reactors—have been adequately addressed by the designed reactor protection systems. In water cooled small modular reactors (SMRs), droplet entrainment in steam flow has significant implications on the progression of accident scenarios due to its compact design features, which requires representative test data applicable to SMR designs. Computer code, modeling and simulation (M&S) tools and models require adequate verification, assessment, and qualification. This includes M&S results validation against scaled empirical data within allowable uncertainty bands to gain regulatory approvals during the various stages of reactor system design, demonstration, and commercialization. However, measurement uncertainty within the empirical datasets and test data applicability ranges requires careful consideration of M&S inputs (i.e., boundary conditions, and initial conditions), and verification and validation efforts. This study focuses on uncertainty quantification in designing scaled test facilities for SMR applications with appropriate measurements and a standard data-reduction method to estimate thermal hydraulics characteristics parameters that incorporate physics phenomena of interest. In addition, this study supports the evaluation model development and assessment process using M&S that interfaces with advanced computing tools and digital twin capabilities. This will allow synchronization between experiment and modeling approaches for droplet entrainment testing and analysis, improving diagnostics, prognostics, and decision-making to accelerate regulatory approval.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗