A feedforward-feedback-based reactor power decoupling control strategy for multi-modular nuclear power plants

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.

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ASME Operation and Maintenance (OM) Code Overview

Volume 1: Operations and Maintenance, Aging Management and Plant Upgrades; Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory; Plant Systems, Structures, Components and Materials; I&C, Digital Controls, and Influence of Human Factors ◽

10.1115/icone24-60086 ◽

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.

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

ASME 2005 Power Conference ◽

10.1115/pwr2005-50003 ◽

2005 ◽

Cited By ~ 1

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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Nuclear Power Plant Fires and Explosions: Part I — Plant Designs and Hydrogen Ignition

Volume 4: Fluid-Structure Interaction ◽

10.1115/pvp2017-66285 ◽

2017 ◽

Cited By ~ 1

Author(s):

Robert A. Leishear

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Nuclear Reactor ◽

Hydrogen Generation ◽

Reactor Power ◽

Pressurized Water ◽

Hydrogen Ignition ◽

History Of ◽

Water Reactors

Requiring further investigation, hydrogen explosions and fires have occurred in several operating nuclear reactor power plants. Major accidents that were affected by hydrogen fires and explosions included Chernobyl, Three Mile Island, and Fukushima Daiichi. Smaller piping explosions have occurred at Hamaoka and Brunsbüttel Nuclear Power Plants. This paper is the first paper in a series of publications to discuss this issue. In particular, the different types of reactors that have a history of fires and explosions are discussed here, along with a discussion of hydrogen generation in commercial reactors, which provides the fuel for fires and explosions in nuclear power plants. Overall, this paper is a review of pertinent information on reactor designs that is of particular importance to this multi-part discussion of hydrogen fires and explosions. Without a review of reactor designs and hydrogen generation, the ensuing technical discussions are inadequately backgrounded. Consequently, the basic designs of pressurized water reactors (PWR’s), boiling water reactors (BWR’s), and pressure-tube graphite reactors (RBMK) are discussed in adequate detail. Of particular interest, the Three Mile Island design for a PWR is presented in some detail.

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Dynamic Testing in Nuclear Power Plants for Model Validation

Journal of Dynamic Systems Measurement and Control ◽

10.1115/1.3427047 ◽

1976 ◽

Vol 98 (3) ◽

pp. 340-347 ◽

Cited By ~ 1

Author(s):

T. W. Kerlin ◽

E. M. Katz ◽

A. T. Chen ◽

J. G. Thakkar ◽

S. I. Chang

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Dynamic Testing ◽

Reactor Power ◽

Normal Operation ◽

Fuel Temperature ◽

Variable Model ◽

Pressurized Water ◽

Frequency Responses

Dynamics tests were preformed at the Oconee pressurized water reactor to obtain information for checking a theoretical plant model. Low level, periodic reactivity perturbations were introduced and several system responses (reactor power, temperatures, pressures) were monitored. The data were processed off-line to give frequency responses. A linear state-variable model for the plant was formulated and used to compute theoretical frequency responses. A computerized, model-reference identification procedure was used to identify the fuel temperature coefficient of reactivity and the overall fuel-to-coolant heat transfer coefficient. The study showed that dynamic tests can be performed in operating nuclear power plants with insignificant interference to normal operation. Also, the use of automatic parameter identification procedures was demonstrated.

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Finding Megawatts in Nuclear Power Plants With Process Data Reconciliation

12th International Conference on Nuclear Engineering, Volume 2 ◽

10.1115/icone12-49152 ◽

2004 ◽

Cited By ~ 2

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.

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ASME OM Code Appendix IV

ASME/NRC 2014 12th Valves, Pumps, and Inservice Testing Symposium ◽

10.1115/nrc2014-5018 ◽

2014 ◽

Author(s):

Fred Setzer

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Reactor Power ◽

Light Water Reactor ◽

Light Water ◽

Water Reactor ◽

Operation And Maintenance ◽

American Society

This paper presents a discussion of the activities ongoing within the ASME (formerly the American Society of Mechanical Engineers) Operation and Maintenance of Nuclear Power Plants (OM) Code Subgroup on Air Operated Valves (SG-AOV), along with an overview of Revision 0 of Mandatory Appendix IV, “Preservice and Inservice Testing of Active Pneumatically Operated Valve Assemblies in Light-Water Reactor Power Plants.” Paper published with permission.

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