Nuclear Regulatory Acceptance of Certified Process Data Reconciliation (CPDR)

2021 ◽  
Author(s):  
Andy Jansky ◽  
Magnus Langenstein
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Plant Performance ◽  
Data Reconciliation ◽  
Operational Parameters ◽  
Process Data ◽  
Power Recovery ◽  
Robust Process ◽  
Substantial Change

Abstract The utility industry is currently undergoing a substantial change from an analogue to a digital infrastructure. Not only plant performance and utility profits are dependent on accurate plant operational parameters but, more importantly, set safety limits need to be met in order to ensure safe operation of nuclear power plants in particular. Using non-quality-assured process data for operational decisions can result in significant over- or under-power of the plant. In addition, all new technologies such as AI, IIoT, digital twin technology, etc. rely on robust process data as input, putting at risk the significance of the results from the continuing data processing ("garbage in, garbage out"). One method, certified process data reconciliation, or CPDR, cuts through the vast amount of available process data and generates all relevant process values with the smallest uncertainty possible. 95% of all collected process data can be discarded after introduction of CPDR. With CPDR, plant operation and maintenance can be significantly optimized and utilities can profit by realizing e.g. power recovery and measurement uncertainty recapture (MUR). Because the focus on reconciled instead of measured values constitutes a paradigm shift, the application of CPDR needs to be communicated to Nuclear Regulators. This paper describes the approach and experience of the Regulator acceptance process in various countries around the globe.

Nuclear Regulatory Acceptance of Certified Process Data Reconciliation (CPDR)

2020 ◽  
Author(s):  
Andy Jansky ◽  
Magnus Langenstein
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Plant Performance ◽  
Data Reconciliation ◽  
Operational Parameters ◽  
Process Data ◽  
Power Recovery ◽  
Robust Process ◽  
Substantial Change

Abstract The utility industry is currently undergoing a substantial change from an analog to a digital infrastructure. Not only plant performance and utility profit are dependent on accurate plant operational parameters but, more importantly, set safety limits need to be met in order to ensure safe operation of nuclear power plants in particular. Using non-quality-assured process data for operational decisions can result in significant over- or under-power of the plant. In addition, all new technologies such as AI, IIoT, digital twin technology, etc. rely on robust process data as input, putting at risk the significance of the results from the continuing data processing (“garbage in, garbage out”). One method, certified process data reconciliation, or CPDR, cuts through the vast amount of available process data and generates all relevant process values with the smallest uncertainty possible. 95% of all collected process data can be discarded after introduction of CPDR. With CPDR, plant operation and maintenance can be significantly optimized and utilities can profit by realizing e.g. power recovery and measurement uncertainty recapture (MUR). Because the focus on reconciled instead of measured values constitutes a paradigm shift, the application of CPDR needs to be communicated to Nuclear Regulators. This paper describes the approach and experience of the Regulator acceptance process in various countries around the globe.

Finding Megawatts in Nuclear Power Plants With Process Data Reconciliation

2004 ◽  
Author(s):  
Magnus Langenstein ◽  
Josef Jansky ◽  
Bernd Laipple ◽  
Horst Eitschberger ◽  
Eberhard Grauf ◽  
...  
Keyword(s):  
Mass Flow ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water Mass ◽  
Reactor Power ◽  
Data Reconciliation ◽  
Thermal Reactor ◽  
Feed Water ◽  
Process Data

Process data reconciliation with VALI III is a method for monitoring and optimising industrial processes as well as for component diagnosis, condition-based maintenance and online calibration of instrumentation. Employing process data reconciliation in nuclear power plants enables thermal reactor power to be determined with an uncertainty of less than ± 0.5%, without having to install additional precision instrumentation to measure as for example the final feed-water mass flow. This is equivalent to a measurement uncertainty recapture power uprate potential of about 1.5% (maximum allowed potential is 2.0%). In addition, process data reconciliation is able to detect any drift in the measured values at an early stage, yet allowing for the reconciled variables (such as thermal reactor power) to be calculated with consistently high precision. Without process data reconciliation • drift in measured values and • systematic errors for the feed-water temperature or the feed-water mass flow could remain undetected. With such measurements the thermal reactor power calculation may incorporate an unacceptably large deviation, which has a negative impact on both, safety and economical aspects. This paper describes, how process data reconciliation works and shows examples of the finding and gain of more than 30 MW electrical power in PWR and BWR units in Germany and Switzerland.

Operating Reliability Assessment of FPGA-Based NPP I&C Systems: Approach, Technique and Implementation

2017 ◽  
Author(s):  
Eugene Babeshko ◽  
Ievgenii Bakhmach ◽  
Vyacheslav Kharchenko ◽  
Eugene Ruchkov ◽  
Oleksandr Siora
Keyword(s):  
Integrated Circuits ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Modes ◽  
New Technologies ◽  
Systems Approach ◽  
Reliability Assessment ◽  
Design Stage ◽  
Fpga Design ◽  
Operating Reliability

Operating reliability assessment of instrumentation and control systems (I&Cs) is always one of the most important activities, especially for critical domains like nuclear power plants (NPPs). Intensive use of relatively new technologies like field programmable gate arrays (FPGAs) in I&C which appear in upgrades and in newly built NPPs makes task to develop and validate advanced operating reliability assessment methods that consider specific technology features very topical. Increased integration densities make the reliability of integrated circuits the most crucial point in modern NPP I&Cs. Moreover, FPGAs differ in some significant ways from other integrated circuits: they are shipped as blanks and are very dependent on design configured into them. Furthermore, FPGA design could be changed during planned NPP outage for different reasons. Considering all possible failure modes of FPGA-based NPP I&C at design stage is a quite challenging task. Therefore, operating reliability assessment is one of the most preferable ways to perform comprehensive analysis of FPGA-based NPP I&Cs. This paper summarizes our experience on operating reliability analysis of FPGA based NPP I&Cs.

scholarly journals CS Measures for Nuclear Power Plant Protection: A Systematic Literature Review

Signals ◽  
2021 ◽  
Vol 2 (4) ◽  
pp. 803-819
Author(s):  
Nabin Chowdhury
Keyword(s):  
Literature Review ◽  
Systematic Literature Review ◽  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Plant Protection ◽  
Security Requirements ◽  
Protection Measures ◽  
Defense In Depth

As digital instrumentation in Nuclear Power Plants (NPPs) is becoming increasingly complex, both attack vectors and defensive strategies are evolving based on new technologies and vulnerabilities. Continued efforts have been made to develop a variety of measures for the cyber defense of these infrastructures, which often consist in adapting security measures previously developed for other critical infrastructure sectors according to the requirements of NPPs. That being said, due to the very recent development of these solutions, there is a lack of agreement or standardization when it comes to their adoption at an industrial level. To better understand the state of the art in NPP Cyber-Security (CS) measures, in this work, we conduct a Systematic Literature Review (SLR) to identify scientific papers discussing CS frameworks, standards, guidelines, best practices, and any additional CS protection measures for NPPs. From our literature analysis, it was evidenced that protecting the digital space in NPPs involves three main steps: (i) identification of critical digital assets; (ii) risk assessment and threat analysis; (iii) establishment of measures for NPP protection based on the defense-in-depth model. To ensure the CS protection of these infrastructures, a holistic defense-in-depth approach is suggested in order to avoid excessive granularity and lack of compatibility between different layers of protection. Additional research is needed to ensure that such a model is developed effectively and that it is based on the interdependencies of all security requirements of NPPs.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

2025 Nuclear Code: The Vision for the Future of ASME Nuclear Codes and Standards

2018 ◽  
Author(s):  
Dale E. Matthews ◽  
Ralph S. Hill ◽  
Charles W. Bruny
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Fusion Energy ◽  
Life Cycle Management ◽  
Economic Benefits ◽  
Plant Design ◽  
Element Analysis ◽  
Energy Devices ◽  
The Future

ASME Nuclear Codes and Standards are used worldwide in the construction, inspection, and repair of commercial nuclear power plants. As the industry looks to the future of nuclear power and some of the new plant designs under development, there will be some significant departures from the current light water reactor (LWR) technology. Some examples are gas-cooled and liquid metal-cooled high temperature reactors (HTRs), small modular reactors (SMRs), and fusion energy devices that are currently under development. Many of these designs will have different safety challenges from the current LWR fleet. Variations of the current LWR technology are also expected to remain in use for the foreseeable future. Worldwide, many LWRs are planned or are already under construction. However, technology for construction of these plants has advanced considerably since most of the current construction codes were written. As a result, many modern design and fabrication methods available today, which provide both safety and economic benefits, cannot be fully utilized since they are not addressed by Code rules. For ASME Nuclear Codes and Standards to maintain and enhance their position as the worldwide leader in the nuclear power industry, they will need to be modernized to address these items. Accordingly, the ASME Nuclear Codes and Standards organizations have initiated the “2025 Nuclear Code” initiative. The purpose of this initiative is to modernize all aspects of ASME’s Nuclear Codes and Standards to adopt new technologies in plant design, construction, and life cycle management. Examples include modernized finite element analysis and fatigue rules, and incorporation of probabilistic and risk-informed methodology. This paper will present the vision for the 2025 ASME Nuclear Codes and Standards and will discuss some of the key elements that are being considered.

Instrumentation and Controls Design and Maintenance Strategy for Small Modular Reactors

2014 ◽  
Author(s):  
Alex H. Hashemian ◽  
Hash M. Hashemian ◽  
Tommy C. Thomasson ◽  
Jeffrey R. Kapernick
Keyword(s):  
Response Time ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Technologies ◽  
Early Stage ◽  
Department Of Energy ◽  
Forced Flow ◽  
Small Modular Reactors ◽  
And Control

Small Modular Reactors (SMRs) under design and development today are working to crystallize the measurements that must be made to control the reactor and monitor its safety. Traditionally, temperature, pressure, level, flow, and neutron flux are measured in conventional nuclear reactors for operation and control and to protect against equipment and process deviations that can affect safety. In most current SMR designs, essentially the same process variables may have to be measured; especially primary coolant flow depending on whether the core cooling and heat transfer results from natural circulation or forced flow. The flow can be measured directly or inferred from other measurements or estimated through empirical or physical modeling. The conventional sensors that are qualified for nuclear services and are currently used in nuclear power plants may or may not be suitable for SMRs. It all depends on the size and qualification requirements, installation details, static and dynamic performance specifications, wiring details, and sensor life expectancy. This paper will explore the possibilities that exist for SMRs to use today’s sensors and any need for new sensor designs. In addition, the paper will identify new means for automated monitoring of instrumentation and control (I&C) sensor performance in SMRs. In particular, the existing array of online calibration monitoring techniques and in-situ response time measurement methods will be evaluated for implementation in SMRs. This is important at this early stage as SMRs can easily build provisions in their mechanical, electrical, and I&C designs to accommodate online and automated I&C maintenance. For example, it is envisioned that SMRs will not be performing periodic sensor calibrations using classical hands-on procedures. Rather, SMRs are expected to be equipped with new technologies to verify the I&C performance automatically and flag the sensors and systems to be calibrated, response time tested, repaired, or replaced. The paper will explore these possibilities and will report on a current R&D project that is underway at AMS with funding from the U.S. Department of Energy (DOE) with the goal to adapt the existing online monitoring (OLM) technologies for implementation in SMRs. The existing OLM technologies have been used by AMS in commercial nuclear power plants and research reactors for monitoring of I&C equipment performance including calibration, response time, detection of sensing line blockages, and to distinguish whether a signal anomaly is due to cables/connectors, electromagnetic interference, an end device being a sensor or a pump, other rotating equipment, etc.

scholarly journals New Technologies for Repairing Aging Cables in Nuclear Power Plants: M3LW-14OR0404015 Cable Rejuvenation Report

2014 ◽  
Author(s):  
Kevin L. Simmons ◽  
Leonard S. Fifield ◽  
Matthew P. Westman ◽  
John A. Roberts
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Technologies

Membrane Covered Probes for O2 and H2 Determination - Diffusion through Liquid Films and Solid Membrane in Practice

2021 ◽  
Vol 413 ◽  
pp. 98-105
Author(s):  
Tomas Moucha ◽  
Václav Linek ◽  
Adam Bouřa ◽  
Tomáš Kracík
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
New Technologies ◽  
Hydrogen Diffusion ◽  
Chemical Species ◽  
Hydrogen Sensor ◽  
Liquid Films ◽  
Concentration Measurement ◽  
Process Conditions ◽  
Primary Coolant

In the era of the expansion of hydrogen use, its concentration measurement becomes more important. We further focus on one of the H2 concentration measurement purposes, where the hydrogen diffusion in a solid membrane and in a liquid electrolyte play the key role. To keep optimal process conditions in the primary cooling circuit of nuclear power plants, various chemical species are dosed in. Among the species the concentration of which is monitored in primary coolant, belong oxygen and hydrogen. While plenty of companies offer oxygen sensors suitable for the measurement in the primary coolant, the hydrogen sensor, really selective to H2 concentration, is offered by only one company. It is worth, therefore, accomplishing the development of a hydrogen sensor, which began at UCT Prague in the 1990's and, after several successful measurements in nuclear power plant, interrupted due to fateful events in the research team. We introduce here the results of the first part of contemporary work of the Mass Transfer Laboratory based on new technologies but using the experience from 1990's. Having at disposal modern functional samples to measure both oxygen and hydrogen concentrations, we verified a fair long-term stability of the sensors and, further, we would like to cooperate with an industrial partner to finalize the development of prototypes and start the production of monitoring units.

CONCEPTUAL BIG DATA CASE STUDY TO IDENTIFYING RISKS OF NEW NUCLEAR TECHNOLOGIES

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Jaeheum Yeon ◽  
Mark Czarny ◽  
John Walewski ◽  
Julian Kang
Keyword(s):  
Big Data ◽  
Content Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
New Technologies ◽  
Risk Identification ◽  
Visual Content ◽  
Short Period ◽  
Identification Approach

New technologies associated with nuclear power plants are being introduced regularly. However, many of the risks and uncertainties associated with these new nuclear technologies have yet to be identified. In this study, the risks related to newly-developed nuclear technologies were determined through an extensive review of the extant literature. A qualitative visual content analysis was selected as the research method employed to identify words repeatedly occurring in 147 journal articles. Through this conceptual “big data” approach, frequently mentioned words were identified using a co-occurrence map. The analysis results were then grouped into four categories: fuel resources, operational system designs, nuclear reactor cooling systems, and steam generators. Words used repeatedly to reference these four key categories were determined to also represent potential causes of risk factors. Many texts could be analyzed in a short period of time through the use of visual content analysis software. Frequently emphasized correlating words were then identified. This big data approach can also be applied to additional nuclear power practices to identify other uncertainties. Last, the limitations of a visual content analysis employed as a risk identification approach were revealed through this study.

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