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.

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.

Process Monitoring during Normal Operations at Canadian Nuclear Power Plants

2000 ◽  
Vol 44 (22) ◽  
pp. 811-814 ◽  
Author(s):  
Eric Davey
Keyword(s):  
Process Monitoring ◽  
Observational Studies ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Process Conditions ◽  
Control Room ◽  
Monitoring Process ◽  
Underlying Basis ◽  
Generic Strategy

This paper summarizes the findings from several observational studies to characterize the basis for a process monitoring strategy used by operators in ‘normal’ operations at CANDU nuclear power plants. These studies were undertaken in support of projects to develop improved control room displays and information systems to better support operators in both normal and abnormal operating situations. With the assistance of operators from several plants, an underlying basis for process monitoring was defined and a ‘generic’ strategy for monitoring process conditions in ‘normal’ operations has been established.

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.

Long-term thermal aging performance of welding structures in primary coolant pipes for nuclear power plants

2020 ◽  
Vol 149 ◽  
pp. 107793
Author(s):  
Minyu Fan ◽  
Mingya Chen ◽  
Min Yu ◽  
Wenqing Jia ◽  
Yuanfei Li ◽  
...  
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Primary Coolant ◽  
Aging Performance

Reverse osmosis for the recovery of boric acid from the primary coolant at nuclear power plants

2016 ◽  
Vol 300 ◽  
pp. 107-116 ◽  
Author(s):  
Šárka Bártová ◽  
Pavel Kůs ◽  
Martin Skala ◽  
Kateřina Vonková
Keyword(s):  
Boric Acid ◽  
Reverse Osmosis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Primary Coolant

Environment-Assisted Intergranular Cracking

MRS Bulletin ◽  
1999 ◽  
Vol 24 (7) ◽  
pp. 36-42 ◽  
Author(s):  
J.R. Scully
Keyword(s):  
High Purity ◽  
Nuclear Power ◽  
Power Plants ◽  
Oil And Gas ◽  
Oil And Gas Industry ◽  
Atmospheric Conditions ◽  
Intergranular Cracking ◽  
Intergranular Stress Corrosion ◽  
Ferrous Alloys ◽  
Primary Coolant

Intergranular separation in polycrys-talline materials involves breaking metallic bonds along grain boundaries in response to stress. The surfaces created in this manner expose the grain facets on either side of the original boundary, as shown in Figure 1. This mode of fracture often occurs at much lower fracture stress and energy than cracking by ductile processes through the interior of grains. The exposure of specific materials to certain environments and stress can promote this low-energy, intergranular mode of separation, even when fracture of the same material in vacuum would occur along a ductile transgranu-lar path. Three types of environment-assisted intergranular cracking can occur in a wide variety of alloy/environment systems: intergranular stress-corrosion cracking (IGSCC), intergranular hydrogen embrittlement, and intergranular liquid-metal embrittlement.Figure 1 shows an example of IGSCC. This type of cracking is a pervasive problem in many technological applications, leading to extensive repairs, loss of service function, and safety concerns. IGSCC occurs in the weld-heat-affected zones of stainless-steel pipes in high-purity primary coolant waters within nuclear power plants, and in nickel-based alloys utilized as heat-exchanger tubing when exposed to the high-purity primary as well as secondary coolant waters in power plants. It is also seen in Al-based alloys used for fuselage skins and structural components in military and commercial aircraft when exposed to humid atmospheric conditions. Ferrous alloys used in the oil and gas industry are also susceptible. For instance, IGSCC of mild steels used in buried gas-transmission pipelines is a widespread international problem, leading to explosions when leaking natural gas ignites.

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 Fouling Of Nuclear Steam Generators: Fundamental Studies, Operating Experience and Remedial Measures Using Chemical Additives

2013 ◽  
Vol 2 (1) ◽  
pp. 61-88 ◽  
Author(s):  
C.W. Turner
Keyword(s):  
Economic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Operating Experience ◽  
Feed Water ◽  
Chemical Additives ◽  
Steam Generators ◽  
Primary Coolant ◽  
And Performance

Fouling remains a potentially serious issue that if left unchecked can lead to degradation of the safety and performance of nuclear steam generators (SGs). It has been demonstrated that the majority of the corrosion product transported with the feed water to the SGs accumulates in the SG on the tube-bundle. By increasing the risk of tube failure and acting as a barrier to heat transfer, deposit on the tube bundle has the potential to impair the ability of the SG to perform its two safety-critical roles: provision of a barrier to the release of radioactivity from the reactor coolant and removal of heat from the primary coolant during power operation and under certain post accident scenarios. Thus, it is imperative to develop improved ways to mitigate SG fouling for the long-term safe, reliable and economic performance of nuclear power plants (NPPs). This paper provides an overview of our current understanding of the mechanisms by which deposit accumulates on the secondary side of the SG, how this accumulation affects SG performance and how accumulation of deposit can be mitigated using chemical additives to the secondary heat-transport system. The paper concludes with some key questions that remain to be addressed to further advance our knowledge of deposit accumulation and how it can be controlled to maintain safe, economic performance of nuclear SGs.

scholarly journals Application of reverse osmosis at NPP and verification of the process for primary coolant treatment in temelín nuclear power

2021 ◽  
Vol 7 (2) ◽  
pp. 1-7
Author(s):  
Skala M. ◽  
Kůs P. ◽  
Kotowski J. ◽  
Kořenková H.
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Boric Acid ◽  
Reverse Osmosis ◽  
Nuclear Power ◽  
Power Plants ◽  
Pure Water ◽  
Feed Solution ◽  
Primary Coolant ◽  
Purification System

Drained primary coolant from nuclear power plants containing boric acid is currently treated in the system of evaporators and by ion exchangers. Reverse osmosis as an alternative process to evaporator was investigated. Using reverse osmosis, the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of concentrated boric acid solution together with other components, while permeate stream consists of purified water. In the first phase ofthe project the reverse osmosis modules from several manufactures were tested on a batch laboratory apparatus. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H3BO3 in the retentate stream, separate from the pure water in the permeate stream. Furthermore, the separation capability for other compounds present in primary coolant such as K, Li or NH3 were evaluated. In the final phase of the project the pilot-plant unit of reverse osmosis was tested in nuclear power plant Temelín. It was installed in the Special Purification System SVO-6 for the regeneration of boric acid. The aim of the tests performed in Temelín nuclear power plant was to verify possible use of reverse osmosis for the treatment of primary coolant.

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