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At least 91 records · Page 5

A human automation interaction concept for a small modular reactor control room - 101

Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Use of Data Libraries for IAEA Nuclear Security Assessment Methodologies (NUSAM) [section 5.4]

Data libraries are essential for the characterization of the facility and provide the documented input which enables the facility assessment results and subsequent conclusions. Data Libraries are historical, verifiable, quantified, and applicable collections of testing data on different types of barriers, sensors, cameras, procedures, and/or personnel. Data libraries are developed and maintained as part of any assessment program or process. Data is collected during the initial stages of facility characterization to aid in the model and/or simulation development process. Data library values may also be developed through the use of state testing centers and/or site resources by testing different types of barriers, sensors, cameras, procedures, and/or personnel. If no data exists, subject matter expert opinion and manufacturer's specifications/ testing values can be the basis for initially assigning values, but are generally less reliable and lack appropriate confidence measures. The use of existing data libraries that have been developed by a state testing organization reduces the assessment costs by establishing standard delay, detection and assessment values for use by multiple sites or facilities where common barriers and alarms systems exist.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Improved alarm tracking for better accountability

An alarm system is a vital component of any accelerator, as it provides a warning that some element of the system is not functioning properly. The severity and age of the alarm may sometimes signify whether urgent or deferred attention is required. For example, older alarms may be given a lower priority if an assumption is made that someone else is already investigating it, whereas those of higher severity or alarms that are more current may indicate the need for an immediate response. The alarm history also provides valuable information regarding the functionality of the overall system, thus careful tracking of these data is likely to improve response time, remove uncertainty about the current status and assist in the ability to promptly respond to the same warning/trigger in the future. Since one goal of every alarm display is to be free of alarms, a clear and concise presentation of an alarm along with useful historic annotations can help the end user address the warning more quickly, thus expediting the elimination of such alarm conditions. By defining a discrete set of very specific alarm management states and by utilizing database resources to maintain a complete and easily accessible alarm history, we anticipate facilitated work flow due to more efficient operator response and management of alarms.

43 PARTICLE ACCELERATORS↗

Nominal Device Support for ATCA Shelf Manager

The ATCA standard specifies a mandatory Shelf Manager (ShM) unit which is a key element for the system operation. It includes the Intelligent Platform Management Controller (IPMC) which monitors the system health, retrieves inventory information and controls the Field Replaceable Units (FRUs). These elements enable the intelligent health monitoring, providing high-availability and safety operation, ensuring the correct system operation. For critical systems like ones of tokamak ITER these features are mandatory to support the long pulse operation. The Nominal Device Support (NDS) was designed and developed for the ITER CODAC Core System (CCS), which will be the responsible for plant Instrumentation and Control (I and C), supervising and monitoring on ITER. It generalizes the Enhanced Physics and Industrial Control System (EPICS) device support interface for Data Acquisition (DAQ) and timing devices. However the support for health management features and ATCA ShM are not yet provided. This paper presents the implementation and test of a NDS for the ATCA ShM, using the ITER Fast Plant System Controller (FPSC) prototype environment. This prototype is fully compatible with the ITER CCS and uses the EPICS Channel Access (CA) protocol as the interface with the Plant Operation Network (PON). The implemented solution running in an EPICS Input / Output Controller (IOC) provides Process Variables (PV) to the PON network with the system information. These PVs can be used for control and monitoring by all CA clients, such as EPICS user interface clients and alarm systems. The results are presented, demonstrating the fully integration and the usability of this solution. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Prototype Design, Analysis, and Results for a Liquid Radiological Waste Control Room

Human factors researchers at Idaho National Laboratory are collaborating with a utility partner to address human factors in the modernization of their radiological waste control room. The collaborating utility, Palo Verde, plans to remove all control boards associated with the liquid radiological waste system, which include controls, indicators, and alarm systems, and replace them with modernized digital instrumentation and controls and displays. To be sure the new system either supports current operational performance or enhances it, researchers have carried out three planning and analysis activities: expert operator review, function allocation analysis, and human factor principles and design applications. Researchers utilized access to expert operators to gain information about the system. Operators were asked to detail a procedure exemplifying typical interactions with the liquid radiological waste system and provide it to Idaho National Laboratory with a list of functionalities to incorporate (Section 4.1). Later, operators were asked to provide additional feedback to inform a prototype with adequate realism to support sound analysis of the system design (Section 4.2). Other operators not associated with the utility partner were also involved in experiment design and prototype evaluation (Section 4.2). Due to the extent of the upgrade, care was given to measure the change in the operator’s role (Section 7). As part of the modernization, some plant systems are expected to be equipped with automation or sensors that send information directly to the control where previously operators were required to enter controlled areas to find the information. Such updates made slight changes to operator role requirements, but the impact was negligible. The way the system is operated will remain largely the same, but operator convenience and safety is expected to increase. Additionally, to map operator tasks during the selected procedure, an Operational Sequence Diagram was created (Section 7.1). The diagram helped to account for every interaction and piece of equipment needed to create the prototype design. Alternatively, the diagram can be created following the final design to identify any function reallocations due to redesign. Human factors principles were also applied to the design. An ergonomic assessment was performed on the workspace resulting in various recommendations for placing and designing the new workstation (Section 6). A 3 D mockup of the control room was created to depict the options available with in-depth descriptions of each. Furthermore the prototype design incorporates well known human factors principles. However, to be sure the principles are sound in the control room context, the researches have designed an experiment to measure the impact design decision has on operator performance. The experiment is described in detail in Section 5. Lastly, a tool emerged as this effort began called the NUREG 0711 Crosswalk (Section 8). The tool has multiple uses, but it was primarily intended to communicate how the researcher’s action directly mapped to specific results suggested in the Nuclear Regulatory Commission review guideline. The tool also maps how current results impact future decisions in the modernization process.

47 OTHER INSTRUMENTATION↗

Human Factors and Technical Considerations for a Computerized Operator Support System Prototype

A prototype computerized operator support system (COSS) has been developed in order to demonstrate the concept and provide a test bed for further research. The prototype is based on four underlying elements consisting of a digital alarm system, computer-based procedures, PI&D system representations, and a recommender module for mitigation actions. At this point, the prototype simulates an interface to a sensor validation module and a fault diagnosis module. These two modules will be fully integrated in the next version of the prototype. The initial version of the prototype is now operational at the Idaho National Laboratory using the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Human Systems Simulation Laboratory (HSSL). The HSSL is a full-scope, full-scale glass top simulator capable of simulating existing and future nuclear power plant main control rooms. The COSS is interfaced to the Generic Pressurized Water Reactor (gPWR) simulator with industry-typical control board layouts. The glass top panels display realistic images of the control boards that can be operated by touch gestures. A section of the simulated control board was dedicated to the COSS human-system interface (HSI), which resulted in a seamless integration of the COSS into the normal control room environment. A COSS demonstration scenario has been developed for the prototype involving the Chemical & Volume Control System (CVCS) of the PWR simulator. It involves a primary coolant leak outside of containment that would require tripping the reactor if not mitigated in a very short timeframe. The COSS prototype presents a series of operator screens that provide the needed information and soft controls to successfully mitigate the event.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Study on main control room and human-system interface of generation III in China - 313

This paper focuses on the study on Main Control Room (MCR) and Human-System Interface (HSI) of Generation III in China. An intelligent HSI has been developed to enhance the safety and availability of Nuclear Power Plant (NPP) by improving operational reliability. The key elements of the proposed HSI such as Advanced Alarm System, State Oriented Procedures (SOP), Large Display Panel (LDP), Computerized-base procedures..., that affects what information the operator sees or the system's response to a control input must be empirically evaluated to ensure that the new design does not compromise HSI effectiveness. Based on CPR1000's experience in China, this paper presents the approach used as well as the most relevant aspects of this kind of project for the next generation. This approach will be used in new Generation III nuclear power plant, and also used in modernizing I and C system in currently operative nuclear power plants, in addition to meeting safety requirements and the plant's operational requirements, to improve cost-effective plant and human performance and to reduce likelihood of human errors, to gain maximum benefit of the implemented technology and to increase the performance, resulting in improved plant safety, availability, reliability, and cost-effective operation. The proposed HSI has been validated and demonstrated using the CPR1000 simulator of National Key Laboratory. After sufficient validation, the characteristic design features of the proposed HSI will be reflected in the design of the main control room of the next generation reactors in China. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Design of virtual SCADA simulation system for pressurized water reactor

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Scale 6.1 Enhancements for Nuclear Criticality Safety

The Scale code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used Scale for safety analysis and design. Scale provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. Scale's graphical user interfaces assist with accurate system modeling and convenient access to desired results. Scale 6.1 builds on the existing capabilities and ease-of-use of Scale and provides several new features such as improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The Scale 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to Scale 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SCALE 6: Comprehensive Nuclear Safety Analysis Code System

Version 6 of the Standardized Computer Analyses for Licensing Evaluation (SCALE) computer software system developed at Oak Ridge National Laboratory, released in February 2009, contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package, which is used worldwide. This paper highlights the capabilities of the SCALE system, including continuous-energy flux calculations for processing multigroup problem-dependent cross sections, ENDF/B-VII continuous-energy and multigroup nuclear cross-section data, continuous-energy Monte Carlo criticality safety calculations, Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques, one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations, two- and three-dimensional lattice physics depletion analyses, fast and accurate source terms and decay heat calculations, automated burnup credit analyses with loading curve search, and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

99 GENERAL AND MISCELLANEOUS↗

Enhancements in SCALE 6.1

The SCALE code system developed at Oak Ridge National Laboratory provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a 'plug-and-play' framework with 89 computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.1 builds on the existing capabilities and ease-of-use of SCALE and provides several new features such as enhanced lattice physics capabilities and multigroup Monte Carlo depletion, improved options and capabilities for sensitivity and uncertainty analysis calculations, improved flexibility in shielding and criticality accident alarm system calculations with automated variance reduction, and new options for the definition of group structures for depletion calculations. The SCALE 6.1 development team has focused on improved robustness via substantial additional regression testing and verification for new and existing features, providing improved performance relative to SCALE 6.0, especially in reactor physics calculations and in the nuclear data used for source term characterization and shielding calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

CAS Operator Training Simulator

SAND2022-1525 O The Central Alarm System (CAS) training simulator is a standalone application that will work on any Alarm Control and Display (AC&D) System. Designed to be used by an instructor in a classroom environment, this software plays back scenario alarms and videos usually exported out of Scribe3D or any software that can produce scenario-based videos. Users can select from an array of different scenarios to play. The alarms are sent to an AC&D controller and the videos are sent to a Video Management Server (VMS). Both alarms and associated videos displayed by the intended AC&D system. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

Le, Tam↗

Criticality Calculations with MCNP6 - Practical Lectures

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Lead Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 13, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube, and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc, depositing energy in a Si solid-state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A business case for nuclear plant control room modernization - 352

This paper presents a generic business case for implementation of technology that supports Control Room Modernization (CRM) in existing nuclear power plants (NPPs). The business case is based on advanced digital instrumentation and control (I and C), human-system interface (HSI), and work process technologies that allow NPP operators to conduct control room operations with improved human factors and efficiency, and therefore operate with fewer operational events. Labor savings can be harvested in terms of reduced overtime and redistribution of work. Significant savings are also demonstrated as reduced time to bring the plant back on line after completion of outages. The benefits are quantified to a rough order of magnitude, providing directional guidance to NPPs that are interested in developing similar business cases. The technologies employed in CRM include large overview displays, task-based operator displays, computer-based procedures, advanced alarm systems, computerized operator support systems, and wireless networks, among others. The CRM business case is founded on the following categories of cost savings: - Integrated Computer-Based Procedures for Operators; - Reduction in Corrective Action Program Work Due to Operator Error; - Reduction in Outage Critical Path Time; - Operation of Local Control Panels; - Reduction in Control Room Support; - Reduction in Plant Events Resulting in Generation Loss; - Paperless Control Room Processes. In addition to cost savings, these technologies also offer improvements to indirect ways that have impact on key performance indicators of the plant. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗