Ensemble-based sensitivity analysis of a Best Estimate Thermal Hydraulics model: Application to a Passive Containment Cooling System of an AP1000 Nuclear Power Plant

Non-destructive inspection (NDI) is one of the key elements in ensuring quality of engineering systems and their safe use. This inspection is a very complex task, during which the inspectors have to rely on their sensory, perceptual, cognitive, and motor skills. It requires high vigilance once it is often carried out on large components, over a long period of time, and in hostile environments and restriction of workplace. A successful NDI requires careful planning, choice of appropriate NDI methods and inspection procedures, as well as qualified and trained inspection personnel. A failure of NDI to detect critical defects in safety-related components of nuclear power plants, for instance, may lead to catastrophic consequences for workers, public and environment. Therefore, ensuring that NDI is reliable and capable of detecting all critical defects is of utmost importance. Despite increased use of automation in NDI, human inspectors, and thus human factors, still play an important role in NDI reliability. Human reliability is the probability of humans conducting specific tasks with satisfactory performance. Many techniques are suitable for modeling and analyzing human reliability in NDI of nuclear power plant components, such as FMEA (Failure Modes and Effects Analysis) and THERP (Technique for Human Error Rate Prediction). An example by using qualitative and quantitative assessesments with these two techniques to improve typical NDI of pipe segments of a core cooling system of a nuclear power plant, through acting on human factors issues, is presented.

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Uncertainty and sensitivity analysis of a LBLOCA in a PWR Nuclear Power Plant: Results of the Phase V of the BEMUSE programme

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2011.08.019 ◽

2011 ◽

Vol 241 (10) ◽

pp. 4206-4222 ◽

Cited By ~ 32

Author(s):

M. Perez ◽

F. Reventos ◽

L. Batet ◽

A. Guba ◽

I. Tóth ◽

...

Keyword(s):

Sensitivity Analysis ◽

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Uncertainty And Sensitivity Analysis

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Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

Applied Mechanics and Materials ◽

10.4028/www.scientific.net/amm.479-480.543 ◽

2013 ◽

Vol 479-480 ◽

pp. 543-547

Author(s):

Jong Rong Wang ◽

Hao Tzu Lin ◽

Wan Yun Li ◽

Shao Wen Chen ◽

Chun Kuan Shih

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Cooling System ◽

Safety Issue ◽

Water Injection ◽

Safety Analysis ◽

Spent Fuel ◽

Spent Fuel Pool ◽

Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

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Bratislavia dadayi (Michaelsen 1905) (Annelida, Clitellata, Naididae): discovery of an alien oligochaete in a technogenic fresh water body in Ukraine

Zootaxa ◽

10.11646/zootaxa.4711.2.7 ◽

2019 ◽

Vol 4711 (2) ◽

pp. 349-365

Author(s):

VLADIMIR A. GUSAKOV ◽

ANZHELIKA A. SYLAIEVA

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Fresh Water ◽

Water Body ◽

Nuclear Power ◽

Potential Distribution ◽

Cooling System ◽

European Continent ◽

Previous Record ◽

Sexually Mature

A non-native oligochaete, Bratislavia dadayi (Michaelsen 1905), is recorded from a water body of the cooling system of the Khmelnitsky Nuclear Power Plant (Ukraine). This is the first registration of this species in the central part of the European continent, far from sea and river navigable waterways. The only previous record of B. dadayi in Europe had been from a Belgian estuary. The occurrence in samples taken over several years, and the presence of sexually mature individuals in the Ukrainian population indicate the worm’s successful naturalization in the new habitat. In this paper, we analyze the species’ morphology and abundance in the Ukrainian population and discuss its ecology, current and potential distribution.

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Dynamic Buckling of Axisymmetric Shells

Journal of Applied Mechanics ◽

10.1115/1.3423434 ◽

1974 ◽

Vol 41 (4) ◽

pp. 1063-1068 ◽

Cited By ~ 5

Author(s):

A. Kalnins

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Free Vibration ◽

Dynamic Stability ◽

Nuclear Power ◽

Cooling System ◽

Dynamic Buckling ◽

Reference Surface ◽

Long Duration ◽

Axisymmetric Shells

A procedure for the analysis of dynamic buckling of axisymmetric shells subjected to axisymmetric, periodic loads of long duration is proposed that is based on the calculation of the nonsymmetric modes of free vibration and associated mode integrals over the reference surface of the shell. Numerical results are presented for the evaluation of dynamic stability of an actual shell that is designed for the cooling system of a nuclear power plant.

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Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit

Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors ◽

10.1115/icone2020-16145 ◽

2020 ◽

Author(s):

Shengtao Zhang ◽

Ke Yi

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Heat Sink ◽

Nuclear Power ◽

Cooling System ◽

Water System ◽

Total Loss ◽

Operating Strategy ◽

Service Water ◽

Essential Service

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.

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Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance ◽

10.1115/icone17-75516 ◽

2009 ◽

Author(s):

Arcadii E. Kisselev ◽

Valerii F. Strizhov ◽

Alexander D. Vasiliev ◽

Vladimir I. Nalivayev ◽

Nikolay Ya. Parshin

Keyword(s):

Experimental Data ◽

Nuclear Power ◽

Cooling System ◽

Severe Accident ◽

Numerical Code ◽

Thermal Hydraulics ◽

Vver Fuel ◽

Design Basis ◽

Best Estimate ◽

Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

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