Application of Internal Fire PRA in Elimination of Fire Common Modes

2017 ◽  
Author(s):  
Xiang Liu ◽  
Lihua Huang ◽  
Zeyu Liu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Principle ◽  
Vulnerability Analysis ◽  
Water Pump ◽  
Safety Level ◽  
Physical Separation ◽  
Design Basis ◽  
Safety Function ◽  
Service Water

During the design and construction phases of nuclear power plants (NPPs), fire protection design principle requires that the safety classified equipment within the nuclear island, the service water pump building and galleries should meet the safety goal whenever a design basis fire occurs. The design of passive facilities, such as structure of buildings and physical separation of equipment, should prevent one single fire from failing all redundant equipment of the same safety function. When one single fire fails all redundant equipment of the same function, it is called a fire common mode. By performing fire vulnerability analysis of fire areas or fire zones, analysts can identify the possible fire common modes, and confirm if they are acceptable or not through further analysis. For NPPs in operation, the vulnerability analysis can be performed by examining the details of plant specific cables and equipment. The confirmed unacceptable common modes should be modified or protected through changing cable routing or adding fire coating, etc. Moreover, by employing internal fire probabilistic risk analysis (PRA), analysts can categorize the identified fire common modes according to severity, assess the changes in the risk of internal fire after elimination, and evaluate the benefit of performing elimination, which provides reference for engineering modification by showing the order of priority of each modification. This article introduces the process of identifying fire common modes, performing internal fire PRA and providing suggestions of modification, and illustrates the application of internal fire PRA in eliminating fire common modes and enhancing the safety level of NPPs.

Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

2011 ◽  
Author(s):  
S. Kalyanam ◽  
D.-J. Shim ◽  
P. Krishnaswamy ◽  
Y. Hioe
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Slow Crack Growth ◽  
Nuclear Industry ◽  
Pipe Material ◽  
Service Water ◽  
Hdpe Pipe ◽  
Pipe Materials ◽  
Hdpe Pipes

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

scholarly journals Fracture Analysis and Fatigue Strength Calculation of Anchor Bolt Used in Circulating Water Pump in Nuclear Power Plant

Scanning ◽  
2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Qiang Chen ◽  
Shuai Zu ◽  
Yinhui Che ◽  
Dongxiong Feng ◽  
Yang Li
Keyword(s):  
Fatigue Strength ◽  
Tensile Stress ◽  
Safety Factor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Analysis ◽  
Water Pump ◽  
Axial Vibration ◽  
Circulating Water ◽  
Tool Marks

A circulating water pump is a key equipment of cooling systems in nuclear power plants. Several anchor bolts were broken at the inlet rings of the same type of pumps. The bolts were turned by a special material for seawater corrosion protection. There were obvious turning tool marks at the root of the thread, which was considered as the source of the crack. The fatigue crack extended to the depth of the bolt, causing obvious radiation stripes on the fracture surface, which was a typical fatigue fracture. Obvious overtightening characteristics were found at the head of the broken bolt. Fracture and energy spectrum analysis showed that the bolt was not corroded. The axial vibration of the pump was measured. The static tensile stress along the bolt axis caused by the preload, the axial tensile stress caused by the axial vibration, and the torsional stress were calculated, respectively. According to the fatigue strength theory, the composite safety factor of the bolt fatigue strength was 1.37 when overtightening at 1.2 times the design torque, which was less than the allowable safety factor of 1.5-1.8, so the bolt was not safe, which further verified the conclusion of fracture analysis. The reason for the low safety factor was caused by the overtightening force. The improvement method was to control the bolt preload or increasing the bolt diameter.

Research on Electric and I&C Equipment Safety Function Classification of Nuclear Power Plant

2021 ◽  
Author(s):  
Wang Yuqi ◽  
Sun Qian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Function Analysis ◽  
System Component ◽  
Design Work ◽  
Equipment Safety ◽  
Safety Function

Abstract Classification of System, Component and Structure (SSC) is the base as well as high level demand of nuclear power plant. Equipment classification including electric and Instrument and Control (I&C) equipment is the precondition of correct design regulation and standard. Safety function classification is key pass of electric and I&C equipment classification. This paper researches the method of nuclear power plant electric and I&C equipment safety function classification. Firstly from view of function, it explains the importance of function classification. Then function analysis and classification of equipment is implemented by design order. Lastly from view of accident analysis, function classification is validated, and a complete approach of function classification is formed. The purpose of this paper is the NPP electric and I&C equipment safety function classification as an example, to study and summarize the method of the electric and I&C equipment safety function classification, and to provide the basis for specific items design work according to design requirements. At the same time, a practical method is provided for other similar NPP electric and I&C equipment classification work. The electric and I&C equipment function classification of nuclear power plant satisfy the basic principles requirement of relative nuclear power rules and codes. It provides an important basis of equipment classification for next nuclear power plants.

scholarly journals Seismic Vulnerability Assessment of Site-Vicinity Infrastructure for Supporting the Accident Management of a Nuclear Power Plant

2017 ◽  
Vol 2017 ◽  
pp. 1-7 ◽  
Author(s):  
T. J. Katona ◽  
A. Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Disaster Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Public Utility ◽  
Qualitative Assessment ◽  
Design Basis ◽  
Accident Management ◽  
Plant Site

Nuclear power plants shall be designed to resist the effects of large earthquakes. The design basis earthquake affects large area around the plant site and can cause serious consequences that will affect the logistical support of the emergency actions at the plant, influence the psychological condition of the plant personnel, and determine the workload of the country’s disaster management personnel. In this paper the main qualitative findings of a study are presented that have been performed for the case of a hypothetical 10−4/a probability design basis earthquake for the Paks Nuclear Power Plant, Hungary. The study covers the qualitative assessment of the postearthquake conditions at the settlements around the plant site including quantitative evaluation of the condition of dwellings. The main goal of the recent phase of the study was to identify public utility vulnerabilities that define the outside support conditions of the nuclear power plant accident management. The results of the study can be used for the planning of logistical support of the plant accident management staff. The study also contributes to better understanding of the working conditions of the disaster management services in the region around the nuclear power plant.

The Implementation of the Gradual Movement of Sliding Blocks to a New Method for Evaluating the Seismic Stability of Slopes Reinforced by Geotextile

2018 ◽  
Vol 12 (04) ◽  
pp. 1841010
Author(s):  
Tadashi Kawai ◽  
Makoto Ishimaru
Keyword(s):  
Seismic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Slope Failure ◽  
Failure Process ◽  
Tohoku Earthquake ◽  
Seismic Stability ◽  
Centrifuge Tests ◽  
Stability Of Slopes ◽  
Design Basis

Evaluating the seismic stability of a rock slope typically involves searching for the minimum value of calculated safety factors (SF) for each supposed sliding block. Because only the transient equilibrium is evaluated, the likelihood of any slope failure can be deemed negligible if all the calculated SFs are greater than unity. However, even if some of the calculated SF are less than unity, it cannot be assumed that the slope will collapse. Recently, in the wake of extremely large earthquakes in Japan, the design earthquake standards for nuclear power plants (NPP) have been extended. After the experience of the 2011 off the Pacific coast of Tohoku Earthquake, the designer is expected to consider beyond design basis earthquakes to determine whether more can reasonably be done to reduce the potential for damage, especially where major consequences may ensue [IAEA (2011). IAEA international fact finding expert mission of the Fukushima dai-ichi NPP accident following the Great East Japan Earthquake and Tsunami, Mission report, IAEA]. With this in mind, the method employed to evaluate the seismic performance of the slope surrounding an NPP needs to be capable of doing more than determining the likelihood of failure: it must also consider the process toward failure in the event of an earthquake beyond the design basis. In this paper, a new evaluation flow which considers the failure process is proposed to evaluate the seismic performance of slopes surrounding an NPP. This is followed by confirming the validity of the concepts in the proposed flow chart by re-evaluating centrifuge tests in past literature and the numerical simulations designed for those tests.

High-Temperature Oxidation Modeling of New Perspective Accident Tolerant Fuel Claddings

2019 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
High Temperature ◽  
High Temperature Oxidation ◽  
Power Plants ◽  
Renewable Energy Sources ◽  
Nuclear Industry ◽  
Safety Level ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Design Basis ◽  
New Perspective

Abstract Currently, the comprehension among the specialists and functionaries is getting stronger that the nuclear industry can encounter serious difficulties in development in the case of insufficiently decisive measures to enhance the safety level of nuclear objects. The keen competition with renewable energy sources like wind, solar or geothermal energy takes place presently and is expected to continue in future decades. One of main measures of nuclear safety enhancement could be the drastic renovation of materials used in nuclear industry. The analytical models of high-temperature oxidation of new perspective materials including chromium-nickel-based alloys, zirconium-based cladding with protective chromium coating, FeCrAl alloys and composite claddings on the basis of SiC/SiC in the course of design-basis and beyond-design-basis accidents at nuclear power plants (NPPs) are developed and implemented to severe accident computer running code. The comparison with available experimental data is conducted. The preliminary calculations of nuclear pressurized water reactor loss-of-coolant accidents with new types of claddings demonstrate encouraging results for hydrogen generation rate and integral hydrogen production. It looks optimistic for considerable upgrade of safety level for future generation NPPs using new fuel and cladding materials.

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