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At least 865 records · Page 48

Study of Data-Driven Mesh-Model Optimization in System Thermal-Hydraulic Simulation

Quantification of nuclear power plant safety risk requires a systematic and yet practical approach to identification of accident scenarios, assessment of their likelihood and consequences. Instrumental to this goal is risk-informed safety margin characterization (RISMC) framework, whose realization requires computationally robust and affordable methods for sufficiently accurate simulation of complex multi-dimensional physical phenomena, such as turbulent and multi-phase flow. The CFD-like correlation-based codes with 3D simulation capability (e.g., GOTHIC) ensure computational efficiency using coarse mesh size and the sub-grid phenomena in the boundary layer that can be captured by adequate constitutive correlations. However, the error sources and user effects on the selection of mesh size and models lead to unpredictable simulation error, while rich High-Fidelity (HF) data from experiments and numerical simulation using validated code or DNS are not fully explored. It would be useful to have a “smart” data-driven multi-scale framework in which the low-resolution models can be “taught” to emulate high-resolution models. The objective of this work is to develop and evaluate a physical-based data-driven mesh-model optimization approach (Optimized Mesh/Model Information System, OMIS) to estimate the simulation error and give advice on the optimized selection of coarse mesh size and Low-Fidelity (LF) models for System Thermal Hydraulic (STH) simulation to achieve accuracy comparable to that of HF models. This approach takes advantages of computational efficiency of coarse-mesh simulation and application of Machine Learning (ML) algorithms.

97 MATHEMATICS AND COMPUTING↗

Instrumentation Plan for the OSU High Temperature Test Facility, Revision 4

This instrumentation plan describes the instrumentation requirements and design for the Oregon State University (OSU) High Temperature Test Facility (HTTF). This report outlines HTTF instrumentation design specifications, data acquisition and control system (DACS) design specifications as well as a brief description of the instrumentation to be used in the HTTF. The HTTF is an integral test facility, which has been designed to model the behavior of interest for a Very High Temperature Gas Reactor (VHTR) during the Depressurized Conduction Cooldown (DCC) event. It also has the potential to conduct limited explorations into the progression of the Pressurized Conduction Cooldown (PCC) event and phenomena during normal operations. The facility is scaled 1:4 by height and 1:64 by volume. The maximum core power for the facility is approximately 2.2MW. The core region is of a modular design to allow for both the prismatic block core and the pebble bed core designs to be tested. However, only the prismatic block design is discussed in this report. A correctly scaled integral facility modeling the VHTR that can be used to examine the thermal hydraulic system behavior of the design basis and beyond design basis transients is of significant value. The experimental results obtained in such a program could be used to validate analytical tools and methods being proposed for use with the VHTR. Therefore, it is of importance that the HTTF be instrumented in such a way that the required data is collected to meet its overall goals of phenomena examination and code validation. In addition to the instrumentation used to collect data to achieve the overall goals of the HTTF test program, there is instrumentation in the HTTF, which is primarily used to ensure the safe operation of the test facility. This type of instrumentation shall be termed “Control Instrumentation” and is discussed in this report. This report discusses the DACS, which is a personal computer controlled system that is used to control both the collection and storage of the test matrix and control system data as well as electrical and mechanical operation of the test facility.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Detection rate evaluation of ex-core detectors in the subcritical OPR-1000 reactor

The OPR-1000 is a PWR reactor developed in Korea. One-type ex-core detectors for monitoring of power distributions were installed in the OPR-1000 reactor to alternate the three-types of the ex-core detectors. For the verification of the detection performances, neutron transport calculation was performed by using MCNP5 code. The reaction rate in the ex-core detectors and the neutron flux were evaluated by using MCNP5 code as changing the boron concentration from 1800 ppm to 1122 ppm in the subcritical condition. The reaction rate results in fission chamber show that minimum and maximum values are 0.03577 and 3.33563 reactions/cm{sup 3}-sec, respectively. This study can be directly used for the verification and improvement of fission chamber performance in using one-type ex-core detector. Also, it can be utilized for the production of the reference data in determining neutron source strength. It is expected the proposed simulation method can be utilized to the improvement of the dose monitoring system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Development of a Methodology to Assess the Impact of Non-Conventional Waste Streams Following a Radiological Incident - 18271

The Environmental Protection Agency's (EPA's) Waste Estimation Support Tool (WEST) is a Geographic Information System (GIS)-based decision support tool for estimating waste generated from remediation and cleanup activities following a radiological incident. WEST uses the Federal Emergency Management Agency's (FEMA's) Hazus-MH software's infrastructure information combined with image classification capabilities to estimate the quantity of waste and debris that may result from a radiological incident. WEST can be used to create customized decontamination strategies that allow end-users to evaluate different means of decontamination and how those strategies may impact the resulting waste stream in each geographical locality. This capability can potentially reduce the response/recovery time frame and cost. One of the key observations from past applications of WEST is that planning scenarios are site-specific, especially when dealing with chemical, biological, radiological, or nuclear (CBRN) waste management issues. Furthermore, site-specific issues are not limited to resource limitations or response bottlenecks but should also account for differences in an area's geography and urban footprint, a phenomenon observed during the Fukushima response. For instance, a recent WEST case study found that the amount of waste generated per unit of affected area, for a given decontamination approach, is largely a function of the building surface area in the affected area (relative to the ground surfaces). However, differences in the geometry and density of contaminated objects in the landscape may impact the amount and characteristics of the waste stream. Although buildings typically constitute a large portion of the surface area in urban environments, other objects such as vehicles and trees can be just as prominent. For example, seasonal changes in the amount of foliage present can alter the non-structural surface area calculated for an area like New York City by 20-30%. Very little is known about the impact that these non-conventional waste streams have on overall remediation strategies, or to what extent differences in characteristics of urban versus rural environments, traffic activities, or seasonal variations may have on these factors. This paper will present the proposed methodology to incorporate into WEST the impacts that niche, nonconventional waste streams might have from rural and urban environments following a nuclear power plant incident or detonation of a radiological dispersal device or improvised nuclear device. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

RoboCount{sup TM} 2020 - A Robotic Non-Destructive Waste Assay System - 19331

A robotic system for non-destructive assay of radioactive waste - RoboCount{sup TM} 2020ab - has been designed, manufactured, and a prototype delivered to a customer. RoboCount{sup TM} 2020 is based on an industrial 6-axis articulating robot platform, combined with a dedicated waste package platform and a high-resolution HPGe gamma spectrometer. The system enables measurement of waste packages including drums (100-l, 200-l, 400-l), crates (up to 1200 x 800 x 800 mm), 1 m{sup 3} flexible intermediate bulk containers (FIBC), and others. Assay type can be freely defined, ranging from currently accepted protocols such as segmented gamma scan (SGS) or integral gamma scan (IGS) to new protocols developed for specific non-cylindrical waste packages, e.g., measurement of FIBCs from all four sides and the top. As a result of the robotic hardware flexibility, reconfiguration of the system for a new task is a matter of software changes only. In addition, while primarily geared towards waste characterization, the system can be configured for free release (clearance) by installing a second detector and field-of-view shielding. The prototype is currently in operation at a NPP site in Europe where it is used for D and D waste characterization in crates and FIBCs. Additional features beyond the basic functionality have been implemented such as an automatic lead collimator change based on measured HPGe detector dead time and an automatic liquid nitrogen refill, performed by the robot via a docking station at predefined time intervals. (Use of electrically cooled HPGe detectors is also possible). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Automated Analysis, Reporting, and Archiving for Robotic Nondestructive Assay of Holdup Deposits - 19508

To decommission deactivated gaseous diffusion enrichment facilities, miles of contaminated pipe must be measured. The current method requires thousands of manual measurements, repeated manual data transcription, and months of manual analysis. The Pipe Crawling Activity Measurement System (PCAMS), developed by Carnegie Mellon University and in commissioning for use at the DOE Portsmouth Gaseous Diffusion Enrichment Facility, uses a robot to measure Uranium-235 from inside pipes and automatically log the data. Radiation measurements, as well as imagery, geometric modeling, and precise measurement positioning data are digitally transferred to the PCAMS server. On the server, data can be automatically processed in minutes and summarized for analyst review. Measurement reports are auto-generated with the push of a button. A database specially-configured to hold heterogeneous data such as spectra, images, and robot trajectories serves as archive. This paper outlines the features and design of the PCAMS Post-Processing Software, currently in commissioning for use at the Portsmouth Gaseous Diffusion Enrichment Facility. The analysis process, the analyst interface to the system, and the content of auto-generated reports are each described. Example pipe-interior geometric surface models, illustration of how key report features apply in operational runs, and user feedback are discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Operating displays - new concept: high performance displays - 118

New control rooms in Nuclear Power Plants using digital Distributed Control System (DCS) operated with software displays, are providing thousands of signals to the operating crew. When an event happens, the Flood of information provided to operators could obscure the initiator event increasing the operator response time, work load and complicating the selection of the actions to be performed. There are some initiatives that are trying to minimize the 'avalanche effect' as the alarm engineering filtering and prioritization or diagnostic displays. Likewise, new control rooms are lost the availability to know the plant status just with a quick glance to the panel indicators. Regarding Small Modular Reactors, where in some cases four operators should monitor and control twelve reactors, these problems are even more significant. The challenge is to find a solution, based on current digital systems, that supply information to operators in order to prevent events and take early actions to minimize or avoid any unexpected incident. In order to minimize human errors, Tecnatom has developed a specific set of displays, called 'High Performance Displays' (HPD) based on objectives, work environment limitation, and non-anticipated events. These displays are being design following the HFE standards: applying the NUREG-0711 elements to identify the information that they have to content and the NUREG-0700 to design the interface. The function of these displays is monitoring the plant situation in a way that the operators are able to recognize any potential event at a simple glance but controlling thousands of signals through pattern recognition, symmetries, regular polygons and colors. The objective of the HPD is to provide an integrated function minimizing the tunnel effect and reduce the workload during the diagnosis phase. Moreover, HPD enhance operators to take the right actions to anticipate the response to any event, even avoiding it, if possible This paper analyzes how HPD's have been design as well as presents the results of a preliminary validation of their capabilities. The preliminary validation is a reduce validation in a full scope simulator running two representative scenarios of abnormal and emergency operation. The analysis is focused on the first stage of the event, i.e. the detection phase, HPD's have been installed in a generic full scope simulator of a nuclear power plant, digital controlled. Both scenarios have been performed by two different operator crews. The first operator shift has worked with the conventional human machine interface, that is, displays similar to the ones currently installed in existing nuclear power plant with digital control. The second crew, reproduce the same scenarios but with HPD available. The studio collects preliminary data in order to compare results, draw lesson learned from the experience to improve HPD design, and find solutions in terms of reducing human errors, human workload, and increase the availability of the plant. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Providing plant data analytics through a seamless digital environment - 211

As technology continues to evolve and become more integrated into a worker's daily routine in the commercial nuclear power industry the need for easy access to data becomes a priority. Not only does the need for data increase but the amount of data collected increases. In most cases the data is collected and stored in various software applications, many of which are legacy systems, which do not offer any other option to access the data except through the application's user interface. Furthermore the data get grouped in 'silos' according to work function and not necessarily by subject. Hence, in order to access all the information needed for a particular task or analysis one may have to access multiple applications to gather all the data needed. The industry and the research community have identified the need for a digital architecture and more importantly the need for a Seamless Digital Environment (SDE). An SDE provides a means to access multiple applications, gather the data points needed, conduct the analysis requested, and present the result to the user with minimal or no effort by the user. The study researches the potential approaches to building an analytics solution for equipment reliability, on a small scale, focusing on either a single piece of equipment or a single system. The analytics solution will consist of a data integration layer, predictive and machine learning layer and the user interface layer that will display the output of the analysis in a straight forward, easy to consume manner. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A novel priority selection system for nuclear power plant - 287

The priority selection system is an important component of security level instrument control platform for nuclear power plant, the main function of which is priority processing the manual or automatic command with various security levels and driving engineering safety features and field devices, such as pump and valve. Its stability and safety plays a crucial role in the operation of the power plant. The main objective of the present paper is to describe the new nuclear power plant-oriented priority selection system, which is developed independently by China Nuclear Control System Engineering Company. In this system, the control commands from 1E DCS, ECP and DAS systems are transmitted via hardwire, while those from NC DCS are transmitted by field-bus. The control commands experience priority output to control the field actuators. Meanwhile, in order to control key field devices when failure of the instrument control systems occurs, the issuing interface for native instructions with permissions setting is also offered. The whole system provides collection-feedback functions for field devices, high coverage self-diagnostic functions and periodic test functions for safety systems. Moreover, the feedback information of field devices, the results of self-diagnosis and periodic test will be submitted to DCS of nuclear power plant. In addition, the core part of priority selection system, namely priority selection module, is obtained by pure FPGA technology, which is different from DCS systems and could avoid common cause failure of software effectively. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The problem analysis of data processing performance in nuclear power plant computerized information and control system - 288

This paper is based on the data processing performance problems of operating command delay and system display not updating real-time for nuclear power plant computerized information and control system. The plant computerized information and control system is responsible for monitoring and controlling the plant operation, and it is a key part in the plant DCS system. Its data processing performance will affect the operation stability of nuclear power plant. The architecture, hardware configuration and data communication for plant computerized information and control system among the different levels in DCS are described. The main reasons affecting data processing performance in different signal channels are analyzed according to the plant load test results. The CPU load curves for real time server and central processing server are displayed, although the detailed cause for operating command delay and system display not updating realtime are not found, central processing server load is heavier than real time server which remind us to analyze the data processing of software resource scheduling deeply. So the resource distribution problem of software execution task is also analyzed according to CPU load test results. This paper provides the optimization analysis for data processing performance, and it is available and effective according to the latest test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Online monitoring using cloud-based applications.

Online monitoring using cloud-based applications presentation. Enabling online monitoring using cloud services will require additional sensors, systems and choices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Development of intelligent instruments with embedded HTTP servers for control and data acquisition in a cryogenic setup—The hardware, firmware, and software implementation

The power of Ethernet for control and automation technology is being largely understood by the automation industry in recent times. Ethernet with HTTP (Hypertext Transfer Protocol) is one of the most widely accepted communication standards today. Ethernet is best known for being able to control through internet from anywhere in the globe. The Ethernet interface with built-in on-chip embedded servers ensures global connections for crate-less model of control and data acquisition systems which have several advantages over traditional crate-based control architectures for slow applications. This architecture will completely eliminate the use of any extra PLC (Programmable Logic Controller) or similar control hardware in any automation network as the control functions are firmware coded inside intelligent meters itself. Here, we describe the indigenously built project of a cryogenic control system built for linear accelerator at Inter University Accelerator Centre, known as “CADS,” which stands for “Complete Automation of Distribution System.” CADS deals with complete hardware, firmware, and software implementation of the automated linac cryogenic distribution system using many Ethernet based embedded cryogenic instruments developed in-house. Each instrument works as an intelligent meter called device-server which has the control functions and control loops built inside the firmware itself. Dedicated meters with built-in servers were designed out of ARM (Acorn RISC (Reduced Instruction Set Computer) Machine) and ATMEL processors and COTS (Commercially Off-the-Shelf) SMD (Surface Mount Devices) components, with analog sensor front-end and a digital back-end web server implementing remote procedure call over HTTP for digital control and readout functions. At present, 24 instruments which run 58 embedded servers inside, each specific to a particular type of sensor-actuator combination for closed loop operations, are now deployed and distributed across control LAN (Local Area Network). A group of six categories of such instruments have been identified for all cryogenic applications required for linac operation which were designed to build this medium-scale cryogenic automation setup. These devices have special features like remote rebooters, daughter boards for PIDs (Proportional Integral Derivative), etc., to operate them remotely in radiation areas and also have emergency switches by which each device can be taken to emergency mode temporarily. Finally, all the data are monitored, logged, controlled, and analyzed online at a central control room which has a user-friendly control interface developed using LabVIEW{sup ®}. This paper discusses the overall hardware, firmware, software design, and implementation for the cryogenics setup.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Monte Carlo semi-empirical model for Si(Li) x-ray detector: Differences between nominal and fitted parameters

A detailed characterization of a X-ray Si(Li) detector was performed to obtain the energy dependence of efficiency in the photon energy range of 6.4 - 59.5 keV, which was measured and reproduced by Monte Carlo (MC) simulations. Significant discrepancies between MC and experimental values were found when the manufacturer parameters of the detector were used in the simulation. A complete Computerized Tomography (CT) detector scan allowed to find the correct crystal dimensions and position inside the capsule. The computed efficiencies with the resulting detector model differed with the measured values no more than 10% in most of the energy range.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Computer correction of an image distorted by turbulent atmosphere

The method for computer correction of images distorted by turbulent atmosphere is realised by means of the simplest optical system comprising a telescope and digital TV-camera. Real-time images with the diffraction resolution are obtained at ground paths of length up to 1800 m.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Description of Sensor Assignment Optimization Method as Deployed on a Multi-Node Cluster

Data analytic methods are being developed to address the problem of how to assign a sensor set in a nuclear facility such that a requisite level of process monitoring capability is realized and that the sensor set is sufficiently rich to determine the status of the individual sensors with respect to need for calibration. There is an awareness in the nuclear industry that data analytics combined with rich sensor sets represent a means to improve operations and reduce costs. In the industry the calibration problem has been previously approached as an empirical data-driven problem with several methods having been developed. However, the experience of the utilities over the past ten years with these methods indicates that the absence of physics-based information renders the data-driven approach less reliable. Complicating factors such as the inherent variability of operation (both equipment alignment and operating condition) can confound a pure data-driven approach while there are no rigorous guidelines for determining what constitutes an adequate sensor set. The solution under development to overcome these shortcomings supplements the data analytic method with process information in a so-called process-constrained data-analytic approach. Simple balance equations are written for generic components (e.g., mechanical pump, valve, and heat exchanger). These do not require a priori knowledge of process parameters, such as heat transfer coefficients or friction factors. All that is needed on the part of the utility user is to identify the components and how they are connected. This report describes the development of a parallel computing capability for determining the optimal sensor set. The optimal sensor set problem suffers from the curse of dimensionality. Computation time increases exponentially as the size of the system grows. To overcome this difficulty a pre-conditioner algorithm is developed to find an approximate solution close the actual solution. This serves as a seed for the full-blown algorithm and acts to constrain the space that must searched. The optimization algorithms are described and the implementation on a parallel computing platform is described. The application of the method to a use case we are solving in collaboration with our utility partner served to illustrate how the default sensor set in a nuclear plant may not provide sufficient coverage to infer sensor calibration status.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Surrogate-Based Robust Design for a Non-Smooth Radiation Source Detection Problem

In this paper, we develop and numerically illustrate a robust sensor network design to optimally detect a radiation source in an urban environment. This problem exhibits several challenges: penalty functionals are non-smooth due to the presence of buildings, radiation transport models are often computationally expensive, sensor locations are not limited to a discrete number of points, and source intensity and location responses, based on a fixed number of sensors, are not unique. We consider a radiation source located in a prototypical 250 m × 180 m urban setting. To address the non-smooth properties of the model and computationally expensive simulation codes, we employ a verified surrogate model based on radial basis functions. Using this surrogate, we formulate and solve a robust design problem that is optimal in an average sense for detecting source location and intensity with minimized uncertainty.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗