• Journal of the Korean Society for Railway
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• v.19 no.2
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• pp.145-158
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• 2016

This paper introduces a risk graph which is one method for determining the SIL as a measure of the effectiveness of signaling system. The purpose of this research is to make up for the weakness of the qualitative determination, which has input value ambiguity and a boundary problem in the SIL range. The fuzzy input valuable consists of consequence, exposure, avoidance and demand rate. The fuzzy inference produces forty eight fuzzy rule by adapting the calibrated risk graph in the IEC 61511. The Max-min composition is utilized for the fuzzy inference. The result of the fuzzy inference is the fuzzy value. Therefore, using the de-fuzzification method, the result should be converted to a crisp value that can be utilized for real projects. Ultimately, the safety requirement for hazard is identified by proposing a SIL result with a tolerable hazard rate. For the validation the results of the proposed method, the fuzzy risk graph model is compared with the safety analysis of the signaling system in CENELEC SC 9XA WG A10 report.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.11 no.11
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• pp.4460-4467
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• 2010

Recent advances in computer technology have brought more dependence on software to railway systems and changed to computer systems. Hence, the reliability and safety assurance of the vital software running on the embedded railway system is going to tend toward very critical task. Accordingly, various software test and validation activities are highly recommended in the international standards related railway software. In this paper, we presented an automated analysis tool and standard for software testing coverage in railway system, and presented its result of implementation. We developed the control flow analysis tool estimating test coverage as an important quantitative item for software safety verification in railway software. Also, we proposed judgement standards due to railway S/W Safety Integrity Level(SWSIL) based on analysis of standards in any other field for utilizing developed tool widely at real railway industrial sites. This tool has more advantage of effective measuring various test coverages than other countries, so we can expect railway S/W development and testing technology of real railway industrial sites in Korea.

• Journal of the Korean Society of Safety
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• v.17 no.3
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• pp.36-42
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• 2002

Structural integrity of either a passenger car or a light truck is one of the basic requirements for a full vehicle engineering and development program. The results of the vehicle product performance are measured in terms of ride and handling, durability, noise/vibration/harshness(NVH), crashworthiness and occupant safety. The level of performance of a vehicle directly affects the marketability, profitability and, most importantly, the future of the automobile manufacturer. In this study, we used the virtual proving ground(VPG) approach for obtaining the dynamic characteristics. VPG approach uses a nonlinear, dynamic, finite element code(LS-DYNA3D) which expands the application boundary outside the classic linear, antic assumptions. VPG approach also uses realistic boundary conditions of tire/road surface interactions. To verify the predicted dynamic results, a single lane change test has been performed. The prediction results were compared with the experimental test results, and the feasibility of the integrated CAE analysis methodology was verified.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.918-924
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• 2020

Film boiling is of great importance in nuclear safety as it directly influences the integrity of nuclear fuel in case of accidents involving loss of coolant. Recently, nuclear power plant safety under earthquake conditions has received much attention. However, to the best of our knowledge, there are no existing studies reporting film boiling in an oscillating system. Most previous studies for film boiling were performed on stationary systems. In this study, numerical simulations were performed for saturated film boiling of water on a horizontal surface under low frequencies to investigate the effect of system oscillation on film boiling heat transfer. A coupled level-set and volume-of-fluid method was used to track the interface between the vapor and liquid phases. With a fixed oscillation amplitude, overall, heat transfer decreases with oscillation frequency. However, there is a frequency region in which heat transfer remains nearly constant. This lock-on phenomenon occurs when the oscillation frequency is near the natural bubble release frequency. With a fixed oscillation frequency, heat transfer decreases with oscillation amplitude. With a fixed maximum amplitude of the additional gravity, heat transfer is affected little by the combination of oscillation amplitude and frequency.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2792-2800
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• 2022

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO₂ kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950-1100 ℃. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

• Journal of Veterinary Clinics
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• v.35 no.2
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• pp.39-45
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• 2018

We evaluated the effects of green tea extract (GTE) supplementation at different dilution steps on boar sperm freezing and in vitro fertilization. Sperm intracellular hydrogen peroxide ($H_2O_2$), motility, viability, acrosome integrity and morphology were determined. In addition, sperm IVF parameters (penetration and monospermy) and glutathione (GSH) levels of presumptive zygotes (PZs) were evaluated. Semen was diluted in lactose egg yolk (LEY) and cooled at $5^{\circ}C$ for 3 h (first dilution step) and then diluted in LEY with 9% glycerol and maintained at $5^{\circ}C$ for 30 min (second dilution step). Four experimental groups were compared: first and second dilution steps without GTE (control), first dilution step with GTE (Step 1), second dilution step with GTE (Step 2) and first and second dilution step with GTE (Step 1+2). The spermatozoa were frozen in nitrogen vapor. Higher sperm motility, viability and acrosome integrity after thawing were observed in Step 1, Step 2 and Step 1+2 groups compared with the control (P < 0.05). Lower $H_2O_2$ level was observed in Step 1+2 compared with control and Step 1 (P < 0.05). For IVF, matured oocytes were co-cultured with spermatozoa frozen according to the experimental groups. GSH levels of PZs were significantly higher in Step 2 and Step 1+2 than in control and Step 1 (P < 0.05) without a significant difference in IVF parameters. In conclusion, supplementation with GTE in both first and second dilution steps during the freezing process resulted in better boar sperm cryopreservation and might be beneficial for further embryo development.

• Fire Science and Engineering
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• v.27 no.5
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• pp.26-31
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• 2013

In nuclear power plant, anchor bolts for pressurizer supports are sufficiently used in terms of safety reason, but field inspections have reported that some bolts exceed the limit of their allowable hardness. Because the high level of hardness may lead to failures due to the stress corrosion or fracture toughness, a regular inspection is required for the bolts in nuclear power plant. Thus, this research measures the hardness of bolts currently used in pressurizer supports and then estimates maximum allowable stresses preventing failures by stress corrosion and fracture toughness. Using the ANSYS program, the stresses of the bolts in the regular condition and accidental condition have been calculated, and the possible maximum stress has been compared with the estimated allowable stresses. From the results, the stresses of bolts in the accidental condition satisfy the allowable safety stress from the stress corrosion failure. However, in the future, it shall be needed to consider the reflection of the structure assembling method on the assembling procedure to ensure the pressurizer integrity during maintenance period time.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Journal of Radiation Industry
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• v.18 no.2
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• pp.133-140
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• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.