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.

Global Balance of Plant in NPPs Using Process Data Reconciliation According to VDI 2048

2014 ◽  
Author(s):  
Magnus Langenstein ◽  
Bernd Laipple
Keyword(s):  
Monitoring System ◽  
Mass Flow ◽  
Online Monitoring ◽  
Water Mass ◽  
Data Reconciliation ◽  
Measurement Information ◽  
Feed Water ◽  
Process Data ◽  
Operation Conditions ◽  
Plant Process

The large quantities of measurement information gathered throughout a plant process make the closing of the mass and energy balance nearly impossible without the help of additional tools. For this reason, a variety of plant monitoring tools for closing plant balances was developed. A major problem with the current tools lies in the non-consideration of redundant measurements which are available throughout the entire plant process. The online monitoring reconciliation system is based on the process data reconciliation according to VDI 2048 standard and is using all redundant measurements within the process to close mass and energy balances. As a result, the most realistic process with the lowest uncertainty can be monitored. This system is installed in more than 35 NPPs worldwide and is used ○ as a basis for correction of feed water mass flow and feed water temperature measurements (recover of lost Megawatts). ○ as a basis for correction of Taverage (Tav) (recover of steam generator outlet pressure in PWRs). ○ for maintaining the thermal core power and the feed water mass flow under continuous operation conditions. ○ for automatic detection of erroneous measurements and measurement drift. ○ for detection of inner leakages, non-condensable gases and system losses. ○ for calculating non measured values (e.g. heat transfer coefficients, ΔT, preheater loads,…). ○ as a monitoring system for the main thermodynamic process. ○ for verifying warranty tests more accurate. ○ as a application of condition-based maintenance and component monitoring. ○ for What-if scenarios (simulation, not PDR) This paper describes the methodology according to VDI 2048 (use of Gaussian correction principle and quality criterias). The benefits gained from the use of the online monitoring system are demonstrated.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

scholarly journals A Framework for Monitoring and Fault Diagnosis in Nuclear Power Plants Based on Signed Directed Graph Methods

2021 ◽  
Vol 9 ◽  
Author(s):  
Wu Guohua ◽  
Yuan Diping ◽  
Yin Jiyao ◽  
Xiao Yiqing ◽  
Ji Dongxu
Keyword(s):  
Directed Graph ◽  
Process Monitoring ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Standard ◽  
Fault Detection And Diagnosis ◽  
Radioactive Material ◽  
Feed Water ◽  
Signed Directed Graph

When nuclear power plants (NPPs) are in a state of failure, they may release radioactive material into the environment. The safety of NPPs must thus be maintained at a high standard. Online monitoring and fault detection and diagnosis (FDD) are important in helping NPP operators understand the state of the system and provide online guidance in a timely manner. Here, to mitigate the shortcomings of process monitoring in NPPs, five-level threshold, qualitative trend analysis (QTA), and signed directed graph (SDG) inference are combined to improve the veracity and sensitivity of process monitoring and FDD. First, a three-level threshold is used for process monitoring to ensure the accuracy of an alarm signal, and candidate faults are determined based on SDG backward inference from the alarm parameters. According to the candidate faults, SDG forward inference is applied to obtain candidate parameters. Second, a five-level threshold and QTA are combined to determine the qualitative trend of candidate parameters to be utilized for FDD. Finally, real faults are identified by SDG forward inference on the basis of alarm parameters and the qualitative trend of candidate parameters. To verify the validity of the method, we have conducted simulation experiments, which comprise loss of coolant accident, steam generator tube rupture, loss of feed water, main steam line break, and station black-out. This case study shows that the proposed method is superior to the conventional SDG method and can diagnose faults more quickly and accurately.

ASME Operation and Maintenance (OM) Code Overview

2016 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Power ◽  
Regulatory Requirements ◽  
General Overview ◽  
Operation And Maintenance ◽  
American Society ◽  
General Organization

The purpose of this paper is to provide a general overview of the organization and content of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code. This will involve a brief description of the regulatory requirements associated with Inservice Testing (IST) as well as a brief overview of the OM Code scope and requirements. This paper will discuss, in general, the regulations requiring IST as well as a brief discussion on when Preservice Testing (PST) and IST become required. A general organization of the ASME OM Code will be provided as well as general topics associated with how to determine when testing and examination intervals are established; what documentation is required; and general discussion regarding the various subsections of the OM Code and the components associated with the OM Code. Alternatives to the OM Code requirements and how to obtain these alternatives will also be provided as well as how the edition applicability of the ASME OM Code is determined. There is also discussion regarding a few general issues associated with the OM Code regarding existing reactor power plants as well as the “new builds” and advanced reactor plants and designs.

Power Recapture and Power Uprate in NPPs With Process Data Reconciliation in Accordance With VDI 2048

2006 ◽  
Author(s):  
Magnus Langenstein
Keyword(s):  
Process Parameters ◽  
Control Volume ◽  
Reactor Power ◽  
Data Reconciliation ◽  
Thermal Reactor ◽  
Process Data ◽  
Pressurized Water ◽  
Statistical Probability ◽  
Water Reactor ◽  
Heat Cycle

The determination of the thermal reactor power is traditionally done by establishing the heat balance: • for a boiling water reactor (BWR) at the interface of reactor control volume and heat cycle; • for a pressurized water reactor (PWR) at the interface of the steam generator control volume and turbine island on the secondary side. The uncertainty of these traditional methods is not easy to determine and it can be in the range of several percent. Technical and legal regulations (e.g. 10CFR50) cover an estimated instrumentation error of up to 2% by increasing the design thermal reactor power for emergency analysis to 102% of the licensed thermal reactor power. Basically, the licensee has the duty to warrant at any time operation inside the analysed region for thermal reactor power. This is normally done by keeping the indicated reactor power at the licensed 100% value. A better way is to use a method which allows a continuous warranty evaluation. The quantification of the level of fulfilment of this warranty is only achievable by a method which: • is independent of single measurements accuracies; • results in a certified quality of single process values and for the total heat cycle analysis; • leads to complete results including 2-sigma deviation especially for thermal reactor power. This method, which is called ‘process data reconciliation based on VDI 2048 guideline’, is presented here [1, 2]. The method allows to determine the true process parameters with a statistical probability of 95%, by considering closed material, mass- and energy balances following the Gaussian correction principle. The amount of redundant process information and complexity of the process improves the final results. This represents the most probable state of the process with minimized uncertainty according to VDI 2048. Hence, calibration and control of the thermal reactor power are possible with low effort but high accuracy and independent of single measurement accuracies. Furthermore, VDI 2048 describes the quality control of important process parameters. Applied to the thermal reactor power, the statistical certainty of warranting the allowable value can be quantified. This quantification allows keeping a safety margin in agreement with the authority. This paper presents the operational application of this method at an operating plant and describes the additional use of process data reconciliation for acceptance tests, power recapture and system and component diagnosis.

scholarly journals A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

2021 ◽  
Vol 7 (4) ◽  
pp. 26-33
Author(s):  
Quang Huy Pham ◽  
Sang Yong Lee ◽  
Seung Jong Oh
Keyword(s):  
Coping Strategies ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Water Injection ◽  
Cooling Water ◽  
Feed Water ◽  
Test Loop ◽  
Station Blackout

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

Warranty Test for a High-Pressure Turbine Retrofit Using Process Data Reconciliation in Accordance With VDI 2048

2008 ◽  
Author(s):  
Magnus Langenstein ◽  
Steffen Riehm ◽  
Jan Hansen-Schmidt
Keyword(s):  
High Pressure ◽  
Statistical Approach ◽  
Mathematical Statistic ◽  
Reactor Power ◽  
Data Reconciliation ◽  
Thermal Reactor ◽  
Process Data ◽  
High Pressure Turbine ◽  
Process Information ◽  
Statistic Approach

The mathematical and statistical approach called “process data reconciliation in accordance with VDI 2048 [1, 2]” • calculates thermal reactor power; • performs warranty tests with on-site operational instrumentation only (without additional instrumentation). with an accuracy of approximately 0.4%. Process data reconciliation based on VDI 2048 is a mathematical-statistic approach that makes use of redundant process information. The overall process monitored continuously in this manner therefore provides hourly process information of a quality equal to acceptance measurements [3 – 10]. At a NPP in Germany a warranty test for a high pressure turbine retrofit was performed in 2007 exclusively using this method. In this paper, the approach and the results of the warranty test are described.

Performance Considerations in Power Uprates of Nuclear Power Plants: A Case Study

2005 ◽  
Author(s):  
Komandur S. Sunder Raj
Keyword(s):  
Flow Measurement ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Power ◽  
Turbine Generator ◽  
Plant Equipment ◽  
The Impact ◽  
Extended Power

This paper examines the impact of power uprates on the performance of nuclear power plants. Since the 1970’s, power companies have been using power uprates to increase the output of their nuclear power plants. The plant systems and components should be capable of accommodating the accompanying increases in flow conditions. The affected components include the turbine-generator, pipes, valves, pumps, heat exchangers, electrical transformer, etc. The Nuclear Regulatory Commission has classified power uprates as falling into three categories: (1) measurement uncertainty recapture power uprates, (2) stretch power uprates and, (3) extended power uprates. Measurement uncertainty recapture power uprates are up to 2% and are achieved by using enhanced techniques for calculating reactor power. This involves the use of state-of-the-art feedwater flow measurement devices to reduce the degree of uncertainty associated with feedwater flow measurement which, in turn, provide for a more accurate calculation of reactor power. Stretch power uprates are typically up to 7% and within the design capacity of the plant. The actual percentage increase in power is plant-specific and depends on the operating margins included in the plant design. Stretch power uprates usually involve changes to instrumentation setpoints, but do not involve major plant modifications. This is especially true for boiling-water reactor plants. In some limited cases where plant equipment is operated at near capacity prior to the power uprate, more substantial changes may be required. Extended power uprates may be up to 20% and, usually require significant modifications to major pieces of plant equipment such as the high pressure turbines, condensate pumps and motors, main generators, and/or transformers. Using a case study, this paper examines the performance considerations involved in power uprates of nuclear power plants. Affected components such as the turbine-generator, moisture separators, reheaters, feedwater heaters and, condensers are discussed. The use of a performance modeling tool in evaluating the impact of power uprates on nuclear plant performance is discussed. The paper provides conclusions and recommendations for ensuring optimal performance in light of power uprates.

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