Deterministic & Probabilistic Evaluations of Structures & Components Credited for Seismic Design Extension Conditions

2021 ◽  
Author(s):  
Ayman Saudy ◽  
Medhat Elgohary
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Method ◽  
Probabilistic Method ◽  
Probabilistic Methods ◽  
Fragility Functions ◽  
Design Features ◽  
High Confidence ◽  
Design Modification ◽  
Electrical Components

Abstract There is "high confidence" in the ability of structures, systems and components (SSCs) of Nuclear Power Plants (NPPs) to perform as designed during Design Basis Accidents. For Design Extension Conditions (DECs), the SSCs are required to perform as designed with "reasonably high confidence." A deterministic design method is proposed to address DECs' higher demands in new and existing CANDU NPPs. The deterministic method builds on the current requirements of applicable codes and standards and recommends more relaxed acceptance criteria. Nevertheless, a means to probabilistically evaluate built-in margins exceeding demand induced by a DEC would provide a measure of the confidence in a DEC-assigned structure or component performing its function. Therefore, a probabilistic method that estimates the probability of survivability for a structure or component when subjected to the demand induced by a DEC is proposed. The probabilistic method could be used to indicate whether there is a need for applying design modification to existing design features to address demands of seismic DEC. The mean, 5-percentile, and 95-percentile fragility functions of these SSCs are used. These fragility functions are typically developed to determine the High-Confidence-Low-Probability-of-Failure value associated with the contribution of a structure or component to the overall plant seismic risk. Sample cases for design features that were implemented in existing CANDU NPPs to address DECs are presented. Both the deterministic and probabilistic methods are applied to cases of Civil structures, passive Mechanical & Electrical components as well as active Control & Instrumentation components.

Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant

2006 ◽  
Author(s):  
Sang-Nyung Kim ◽  
Sang-Gyu Lim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Measurement Data ◽  
Significant Loss ◽  
Design Modification ◽  
Root Cause ◽  
Explosive Expansion ◽  
The Impact

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.

Qualification of Motor Actuated Valve Assemblies for Safety Related Nuclear Applications: Lessons Learned

2018 ◽  
Author(s):  
Ronald Farrell ◽  
L. Ike Ezekoye
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Decision Processes ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
Missed Opportunities ◽  
Ieee Standards ◽  
Electrical Components ◽  
Nuclear Applications

Safety related valves in nuclear power plants are required to be qualified in accordance with the ASME QME-1 standard. This standard describes the requirements and the processes for qualifying active mechanical equipment that are used in nuclear power plants. It does not cover the qualification of electrical components that are addressed using IEEE standards; however, QME-1 recognizes that both mechanical and electrical components must be qualified when they are interfaced as an assembly. Qualifying both mechanical and electrical valve assemblies can be challenging. Considerable amount of judgment is used when developing the plan to qualify any valve with an electric motor actuator. If the wrong steps are taken in planning the tests, the results from the tests may not be useful thus triggering the need to perform additional tests to comply with QME-1 requirements. This paper presents lessons learned in the process of qualifying valve assemblies to meet QME-1 requirements. The lessons include the decision processes associated with planning and executing valve testing, analysis of the valve assemblies for natural frequency determination, and missed opportunities to capture relevant test data during the tests. Finally, the paper will discuss challenges associated with justifying the tests and extending the results of the tests to cover untested valve assemblies.

Seismic Design of Free Standing Racks in Japanese Nuclear Power Plants

2017 ◽  
Author(s):  
Yu Takaki ◽  
Katsuhiko Taniguchi ◽  
Junichi Kishimoto ◽  
Akihisa Iwasaki ◽  
Yoshitsugu Nekomoto ◽  
...  
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Method ◽  
Seismic Resistance ◽  
Spent Fuel ◽  
Analysis Model ◽  
Free Standing ◽  
Scale Free ◽  
Seismic Design Method

The free standing racks are spent fuel storage racks with self-sustained structure without fixation to the pit floor or pit walls. If a free standing rack receives a force to move it due to an earthquake, the force acting on each member of the rack is reduced in compared to the floor-anchored racks owing to sliding of the free standing rack. Now it is planned to exchange the existing floor-anchored racks with the free standing racks to secure higher seismic resistance. In previous studies, efforts were made to establish a behavior analysis model that allows for evaluation of sliding and rocking behaviors of free standing racks and to make out a seismic design method based on an evaluation technique to evaluate, in a conservative manner, vibration test results of full-scale free standing racks. The free standing racks which consist of connected eight racks are designed with this seismic design method. It was confirmed that the free standing racks have enough seismic resistance by performing evaluation using the basic seismic motion and making an analysis on beyond the design event.

Regulatory Aspects on the Application of the Leak Before Break Concept for Swedish Nuclear Power Plants

2008 ◽  
Author(s):  
B. Brickstad
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Probabilistic Methods ◽  
Local Dynamic ◽  
Pipe Failure ◽  
Critical Crack ◽  
Critical Crack Length ◽  
Regulatory Aspects ◽  
Pipe Break ◽  
Leak Before Break

The SKI regulation SKIFS 2004:2 allows for the use of Leak Before Break (LBB) as one way to provide assurance that adequate protection exists against the local dynamic consequences of a pipe break. The way to demonstrate that LBB prevails relies on a deterministic procedure for which a leakage crack is postulated in certain sections of the pipe based on the leak detection capability of the plant. It shall then be demonstrated that certain margins exist against the critical crack length at which a pipe break can be expected. In certain situations, probabilistic methods can strengthen the conclusion that LBB prevails. Then it is necessary to demonstrate that the likelihood of a pipe failure is sufficiently low and that there is a sufficient margin between a detectable leak and pipe rupture. This paper provides some general regulatory aspects on the application of LBB as well as results from two recently completed research projects. The first one is investigating the influence on the deterministic LBB-margins from different assumptions of the crack morphology. The second project provides information on failure probabilities for both leak and rupture for pipes of different sizes in Swedish PWR- and BWR-plants. The interest is focused on the conditional rupture probability corresponding to when the deterministic LBB-margins are fulfilled. Finally some results of failure probability for pipes with stress corrosion cracking are discussed.

scholarly journals Cyber Resilience Analysis of SCADA Systems in Nuclear Power Plants

2020 ◽  
Author(s):  
Meghan Galiardi ◽  
Amanda Gonzales ◽  
Jamie Thorpe ◽  
Eric Vugrin ◽  
Raymond Fasano ◽  
...  
Keyword(s):  
Control System ◽  
Control Systems ◽  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cyber Attacks ◽  
Use Case ◽  
Design Features ◽  
Scada Systems

Abstract Aging plants, efficiency goals, and safety needs are driving increased digitalization in nuclear power plants (NPP). Security has always been a key design consideration for NPP architectures, but increased digitalization and the emergence of malware such as Stuxnet, CRASHOVERRIDE, and TRITON that specifically target industrial control systems have heightened concerns about the susceptibility of NPPs to cyber attacks. The cyber security community has come to realize the impossibility of guaranteeing the security of these plants with 100% certainty, so demand for including resilience in NPP architectures is increasing. Whereas cyber security design features often focus on preventing access by cyber threats and ensuring confidentiality, integrity, and availability (CIA) of control systems, cyber resilience design features complement security features by limiting damage, enabling continued operations, and facilitating a rapid recovery from the attack in the event control systems are compromised. This paper introduces the REsilience VeRification UNit (RevRun) toolset, a software platform that was prototyped to support cyber resilience analysis of NPP architectures. Researchers at Sandia National Laboratories have recently developed models of NPP control and SCADA systems using the SCEPTRE platform. SCEPTRE integrates simulation, virtual hardware, software, and actual hardware to model the operation of cyber-physical systems. RevRun can be used to extract data from SCEPTRE experiments and to process that data to produce quantitative resilience metrics of the NPP architecture modeled in SCEPTRE. This paper details how RevRun calculates these metrics in a customizable, repeatable, and automated fashion that limits the burden placed upon the analyst. This paper describes RevRun’s application and use in the context of a hypothetical attack on an NPP control system. The use case specifies the control system and a series of attacks and explores the resilience of the system to the attacks. The use case further shows how to configure RevRun to run experiments, how resilience metrics are calculated, and how the resilience metrics and RevRun tool can be used to conduct the related resilience analysis.

Democratization of 3D Applications in Nuclear Power Plants: Laser Scanning 3D Technology Implementation on Spanish PWRs

2010 ◽  
Author(s):  
Xavier Jardi´ ◽  
Jorge Anga´s
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Laser Scanning ◽  
Nuclear Power Plants ◽  
Geometrical Parameters ◽  
Radiological Protection ◽  
Web Servers ◽  
3D Technology ◽  
Design Modification ◽  
3D Web

On January 2008, the US NRC issued the Generic Letter 2008-01 [1], “Managing gas accumulation in emergency core cooling, decay heat removal and containment spray systems”. Among other responses, this letter requires an evaluation of locations sensitive to accumulate gases in several safety systems. In order to get accurate data related to the real slope of horizontal pipes and other geometrical parameters needed for this evaluation, laser scanning and 3D modeling techniques have been applied in Spanish Nuclear Power Plants. From October 2008 to December 2009, five Spanish units have been scanned and modeled. As a result of these activities, the plants have obtained detailed 3D models as well as 2D as-built drawings of the selected components. These models were integrated in 3D web servers which give a panoramic view of the scanned areas and permitted measurements in the local coordinate system of the plant. Moreover, the 2D elevation drawings included accurate and useful information for the plants in order to make decisions related to the GL-2008-01 requirements. The geometric information generated in the frame of the GL-2008-01 activities is being currently used for alternative applications. For instance, laser scanning technology is being used to enhance design modification procedures. A pilot project on the MSRs replacement is being currently carried out with successful results. This technology has the advantage that new components from CAD software can be updated in the as-built models obtained through laser scanning. In addition to this, it’s very easy to check fitting and interferences, and also to make accurate measurements and handling simulations. The potential applications in personnel training and radiological protection are also very important. The panoramic viewers on 3D web servers are versatile and could fit the specific requirements of each organization. Regarding staff training, virtual tours and component seekers are being currently developed. These tools provide a significant save of time and dose and also give independence for each person to get to the working place without external help or time-consuming paper consulting. Integration with existing plant databases is also possible through the panoramic viewers and is currently being developed for In-Service Inspections and Maintenance applications. The main advantage of these products is their accessibility with free visors which don’t need specific training. Therefore, the implementation of these tools doesn’t need additional investments. In conclusion, Laser 3D Technology Applications set the first step on the democratization of these powerful 3D environments among common users as integrated tools in their daily work.

Conceptual Formation of Advanced HIS Design Method for Digital I&C + HMIT

2017 ◽  
Author(s):  
Hidekazu Yoshikawa ◽  
Zhanguo Ma ◽  
Amjad Nawaz ◽  
Ming Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Design Method ◽  
Severe Accident ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Human Interface ◽  
Conceptual Frame ◽  
Loss Of Coolant Accident

A new conceptual frame of how to design and validate a digital HIS (human interface system) on an innovative numerical simulation basis is proposed for the support of plant operators’ supervisory control of various types of automated complex NPPs (nuclear power plants). The proposed conceptual framework utilizes the object-oriented AI softwares for plant DiD (defense-in depth) risk monitor with the combination of nuclear reactor accident simulation by an advanced nuclear safety analysis code RELAP5/MOD4 and severe accident analysis code MAAP. The developed conceptual frame proposed in this paper will be applied for an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000.

scholarly journals Risk-Informed Approach to Safety Operation of Units of Nuclear Power Plants

10.12737/189 ◽  
2013 ◽  
Vol 2 (1) ◽  
pp. 14-19
Author(s):  
Ершов ◽  
G. Ershov ◽  
Антонов ◽  
Aleksandr Antonov ◽  
Морозова ◽  
...  
Keyword(s):  
Decision Making ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Probabilistic Methods ◽  
Decision Making Process ◽  
Widespread Application ◽  
Safety Assurance ◽  
Safety Operation ◽  
Method Approach

Safety assurance of NPP units operation is the top priority objective for operating and regulatory organizations. For the purpose of NPP safety assurance deterministic and probabilistic methods the main advantages of which are combined into risk-informed method (approach) are traditionally utilized. Nowadays the risk-informed approach has a widespread application in the riskinformed decision-making process.

Innovation structure combining inter-story isolation with passive cooling effect for AP1000 nuclear power plants

2021 ◽  
Vol ahead-of-print (ahead-of-print) ◽  
Author(s):  
Gangling Hou ◽  
Yu Liu ◽  
Tao Wang ◽  
Binsheng Wang ◽  
Tianshu Song ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Design Method ◽  
Passive Cooling ◽  
Original Design ◽  
Seismic Safety ◽  
Lower Altitude ◽  
Content Type

PurposeAn inter-story isolation structure (IIS) for AP1000 nuclear power plants (NPPs) is provided to resolve the conflict of seismic safety and the optimal location of air intakes.Design/methodology/approachThe effect of passive cooling system (PCS) is better with lower altitude of air intakes than that in the original design of AP1000 NPPs. Seismic performances of IIS NPPs, including the seismic responses, damping frequency bandwidth and seismic reduction robustness, are improved by combining the position of air intakes lower and the optimal design method.FindingsTheoretical analysis and numerical simulation are illustrated that the seismic reduction failure of IIS NPPs is the lowest probability of occurrence when PCS has highest working efficiency.Originality/valueThe IIS NPPs can transfer the contradiction between PCS work efficiency and seismic safety of NPPs to the mutual promotion of them.

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