scholarly journals Quantitative Analysis of Valve Contribution to the Failure Probability of Carbon Dioxide Suppression System

2020 ◽  
Vol 34 (6) ◽  
pp. 79-84
Author(s):  
Hyun-Tae Yim ◽  
Joo-Sung Kim
Keyword(s):  
Carbon Dioxide ◽  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Fire Suppression ◽  
Evaluation Model ◽  
Fire Area ◽  
Residual Air ◽  
Suppression System ◽  
Suppression Systems

Carbon dioxide suppression systems are used in nuclear power plants to extinguish oil fires and ensure integrity of critical equipment. In this study, the contributions of the valves in the carbon dioxide suppression system to the failure probability of suppression were quantitatively analyzed, and the failure probability of the fire suppression system applied to the fire Probabilistic Safety Assessment (PSA) was evaluated for appropriateness. Then, a reliability evaluation model was developed in the form of a fault tree, and the reliability data were analyzed for the major component. The failure probability of the carbon dioxide suppression system with early air-release equipment was assessed as 6.22 × 10<sup>-3</sup>/demand, which is 2.67 times that of the system without such equipment. The results were attributed to the addition of the diverted line isolation valves and quick exhaust valve to release residual air into the atmosphere before supplying carbon dioxide to the fire area. It was also confirmed that the failure probability of the carbon dioxide suppression system with the early air-release equipment was 15.6% that of the suppression provided by the fire PSA and that the failure probability of suppression by the fire PSA was conservative. There are no reported guidelines in literature for analyzing the reliabilities of carbon dioxide fire suppression systems, and the fire PSA currently use the failure probability of suppression recommended by the NSAC-179L.

Application of Supercritical Fluids in Thermal- and Nuclear-Power Engineering

2021 ◽  
pp. 601-658
Author(s):  
Igor L. Pioro
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Supercritical Fluids ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Thermal Power Plants ◽  
Power Cycles

Supercritical Fluids (SCFs) have unique thermophyscial properties and heat-transfer characteristics, which make them very attractive for use in power industry. In this chapter, specifics of thermophysical properties and heat transfer of SCFs such as water, carbon dioxide, and helium are considered and discussed. Also, particularities of heat transfer at Supercritical Pressures (SCPs) are presented, and the most accurate heat-transfer correlations are listed. Supercritical Water (SCW) is widely used as the working fluid in the SCP Rankine “steam”-turbine cycle in fossil-fuel thermal power plants. This increase in thermal efficiency is possible by application of high-temperature reactors and power cycles. Currently, six concepts of Generation-IV reactors are being developed, with coolant outlet temperatures of 500°C~1000°C. SCFs will be used as coolants (helium in GFRs and VHTRs, and SCW in SCWRs) and/or working fluids in power cycles (helium, mixture of nitrogen (80%) and helium (20%), nitrogen and carbon dioxide in Brayton gas-turbine cycles, and SCW/“steam” in Rankine cycle).

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

scholarly journals A Bayesian updating of crack distributions in steam generator tubes

2021 ◽  
Vol 30 (2) ◽  
pp. 33-44
Author(s):  
Alexandre Santos Francisco ◽  
Tiago Simões
Keyword(s):  
Steam Generator ◽  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Plastic Collapse ◽  
Structural Failure ◽  
Maintenance Strategies ◽  
Threshold Size ◽  
Steam Generator Tubes ◽  
Good Agreement

The structural failure of steam generator tubes is a common problem that can a ect the availability and safety of nuclear power plants. To minimize the probability of occurrence of failure, it is needed to implement maintenance strategies such as periodic nondestructive inspections of tubes. Thus, a tube is repaired or plugged whenever it has detected a crack which a threshold size is overtaken. In general, uncertainties and errors in crack sizes are associated with the nondestructive inspections. These uncertainties and errors should be appropriately characterized to estimate the actual crack distribution. This work proposes a Bayesian approach for updating crack distributions, which in turn allows computing the failure probability of steam generator tubes at current and future times. The failure criterion is based on plastic collapse phenomenon, and the failure probability is computed by using the Monte-Carlo simulation. The failure probability at current and future times is in good agreement with the ones presented in the literature.

Transverse microstructure of oxide formed on 20/25/Nb stainless steel

1983 ◽  
Vol 41 ◽  
pp. 214-215
Author(s):  
M.J. Bennett ◽  
J.A. Desport ◽  
Patricia A. Labun
Keyword(s):  
Carbon Dioxide ◽  
Stainless Steel ◽  
Oxide Scale ◽  
Nuclear Power ◽  
Power Plants ◽  
Steel Specimen ◽  
Rolled Sheet ◽  
Fuel Cladding ◽  
The United Kingdom ◽  
Transmission Electron

In the United Kingdom commercial advanced gas cooled nuclear power plants the fuel cladding is a 20% Cr/25% Ni/niobium stabilised (20/25/Nb) austenitic stainless steel and the coolant is carbon dioxide based. The oxidation behaviour of this steel has been studied extensively, e.g., with the main emphasis being directed to the oxidation kinetics and scale composition, as revealed by conventional surface analytical and X-ray diffraction techniques. This paper reports the first transmission electron microscopy undertaken on transverse sections through an oxide scale formed on this steel.The 20/25/Nb steel specimen was a coupon of rolled sheet, which had the composition given in Table. The surface of the steel was abraded on 600 grit SiC paper and subsequently was hydrogen annealed for 30 mins, at 930°C. Oxidation in carbon dioxide was continued for 4900h at 825°C (a probable maximum fuel cladding operating temperature). The magnitudes and kinetics of both oxidation and concurrent spallation have been reported. For the preparation of a TEM foil the oxide scale was protected and the oxide/gas interface was defined by a stainless steel coating deposited by sputter ion plating (SIP).

Mini-Channel Supercritical CO2 Heat Transfer Measurements for Brayton Cycle Regenerators

2009 ◽  
Author(s):  
Alan Kruizenga ◽  
Mark Anderson ◽  
Michael Corradini
Keyword(s):  
Heat Transfer ◽  
Carbon Dioxide ◽  
Critical Point ◽  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Effective Length ◽  
Transfer Coefficients ◽  
Capital Costs ◽  
Mini Channels

Recently, it has become increasingly important to improve efficiency and reduce capital costs in nuclear power plants. This prompted significant work in studying advanced Brayton cycles for high temperature energy conversion. A particular improvement in the operation of an advanced carbon dioxide cycle, is the use of compact, highly efficient, diffusion bonded heat exchangers for the recuperators. These heat exchangers operate near the pseudo-critical point of liquid carbon dioxide, making use of the drastic variation of the thermo-physical properties. This paper focuses on the experimental measurements of heat transfer and pressure drop characteristics within mini-channels. Two test section channel geometries are studied: a straight channel and a zig-zag channel. Both configurations are 0.5m in length and constructed out of 316 stainless steel with a series of nine parallel 1.9mm semi-circular channels. The zig-zag configuration has an angle of 115 degrees with an effective length of ∼0.58m. Heat transfer measurements are conducted for varying ranges of inlet temperatures, pressures, and mass flow rates. Local and average heat transfer coefficients near the critical point are determined from measured wall temperatures and calculated local bulk temperatures.

Failure Probability of Degraded Pipes Based on Probabilistic Fracture Mechanics for Seismic Safety Margin Assessment on NPP

2010 ◽  
Author(s):  
Yinsheng Li ◽  
Masaki Nakagawa ◽  
Katsumi Ebisawa ◽  
Shinobu Yoshimura ◽  
Hiroyuki Kameda
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ground Motions ◽  
Safety Margin ◽  
Mechanical Analysis ◽  
Margin Assessment ◽  
Seismic Safety ◽  
Design Values

The Niigata-ken Chuetsu-oki Earthquake happened in July 2007. Although the observed seismic ground motions of Kashiwazaki-Kariwa nuclear power plants greatly exceeded their design values, the important functions of the plants such as “shutdown”, “cooling”, and “confinement” were successfully ensured. Therefore, assessment of the seismic safety margin of nuclear power plants, both their systems and their components, becomes a very important issue. In this paper, failure probability of cracked pipes in existing nuclear power plants is analyzed by employing an improved probabilistic fracture mechanical analysis code and considering stress corrosion cracking and fatigue. Based on the analysis results of the failure probability and the definitions of the seismic safety margin, the seismic safety margin of degraded pipes in existing nuclear power plants is investigated.

Sign in / Sign up

Export Citation Format

Share Document